ML20244D615
| ML20244D615 | |
| Person / Time | |
|---|---|
| Site: | Oconee, 05000000 |
| Issue date: | 01/07/1986 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Sheron B Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20195F761 | List:
|
| References | |
| FOIA-87-714 GL-81-14, NUDOCS 8601090673 | |
| Download: ML20244D615 (2) | |
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- M%q'o UNITED STATES
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- JAN 07 M
..... 1 Brian Sheron, Deputy Director, Division of Safety Review and MEMORANDUM FOR:
Oversight !
Dennis M. Crutchfield, Assistant Director for Technical FROM:
Support, Division of PWR Licensing-B i
SUBJECT:
REGULATORY ANALYSIS FOR PLANT SPECIFIC BACKFIT OF O The purpose of this memorandum is to requf. the Division of Safety Review !
and Oversight assistance in developing a regulatory ana 1, 2 and 3. Specifically, as a result of staff review of licensee reponses to Generic Letter 81-14 (MPA C-14) concerning This is caused seismic q
ln a loss of all auxiliary feedwater to the three Oconee units. I by the flooding of the turbine building which When this eventresults is compounded whenbythe a nonseismic ci culating water system piping ruptures.
single active failure as is standard licensing practice, the y would exist. Our proposed backfit requires the licensee to correct this situation.
We have recently discussed this subject with J. Rosenthal and E. Chellfah of RRAB and the/ have indicated that the Oconee PRA results can be ut prepare this regulatory analysis.an initial cost / benefitIf be a performed to e a backfit that can be hstified within the recomended guide and ask that a more detailed cost / benefit be developed.
We are also enclosing the We are available to assist in this effort. memorandum which contains pertinent December 16, 1985 this issue. Your prompt attention is requested. i 4 ch ie d, sisantdirector Denn i .
for Technical Suppor Division of PWR Licensing-B f
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Enclosure:
As Stated .,
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Contact:
J. Wermiel ,f 1 ,..-
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K. Kniel F. Congel J. Rosenthal E. Challiah cc w/o enclosure:
F. Miraglia F. Shroeder T. Speis
- 0. Parr J. Stolz R. Weller J. Wermiel H. Nicholaras i
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[pna m o UNITED STATES
!jy#([g p, NUCLEAR REGULATORY COMMISSION, I '~* ?
E WASHINGTON,0. C. 20555 e
....+ gg,1 J 1985 MEMORANDUM FOR: Frank Miraglia, Director, Division of PWR Licensing-B FROM; Dennis Crutchfield, Assistant Director for Technical Support Division of PWR Licensing-B
SUBJECT:
SEISMIC QUALIFICATION OF THE OCONEE AUXILIARY FEEDWATER SYSTEM, MPA C-14 (TAC NOS. 43643/4/5)
The. purpose of this memorandum is to request your approval to proceed with a plant specific backfit of Oconee Units 1, 2 and 3 in order to resolve the remaining key issue with regard to auxiliary feedwater system (AFWS) seismic qualification. Oconee is the only plant remaining under MPA C-14 for which the staff review is not complete. The staff has been negotiating this issue with Duke Power for over three years.
By memorandum from L. S. Rubenstein to Gus C. Lainas dated October 22, 1985 (Enclosure 1), DSI transmitted the safety evaluation report concerning AFWS seismic qualification for Oconee in response to Generic Letter 81-14 (MPA C-14). That SER identified three open items which required resolution in order to demonstrate suitable AFWS seismic qualification. It was also stated in that memorandum that any modifications required at Oconee to resolve the open items were not considered to be backfits according to.Oraft Manual Chapter 0514 based on the fact that Generic Letter 81-14 required conformance with GDC 2 and 34 which are existing staff requirements.
At the request of J. Stolz, we have rerevieweo the SER and DSI findings regarding backfit in order to assist him in developing a course of action for expediting resolution of this issue. We have identified the primary concern regarding AFWS seismic inadequacy to be as follows.
In the event of a safe shutdown earthquake (SSE), because the circulating water system is not seismically supported, it is assumed to rupture and flood the lower elevation of the turbine building which is common to the three units.
Becauss the AFWS pumps for all three Oconee units are located in a comon area at this elevation of the turbine building, they will flood and would be unavailable for reactor shutdown. The licensec has performed an analysis which demonstrates that the tutbine building will maintain its structural integrity in the event of an SSE. The safe shutdown facility (SSF) at Oconee is qualified to withstand an SSE but contains a single seismically qualified AFWS pump. Therefore, a concurrent single active failure will result in no AFWS function for post-earthquake decay heat removal as specified in GDC 34
Contact:
J. Wermiel X29462 4L di M gg
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~2' .. '4 Sff Based on our discussions with the licensee and previous site visits, we believe this issue can be resolved by installation O' seismically supported waterproof walls around the three turbine driven AFWS puups. This modification in con-junction with the SSF would assure et least one AFWS pump for shutdown of the three units in the event of an SSE and concurrent single active failure.
Alternatively, the licenseo could add a second seismically qualified pump to the SSF with sufficient capacity to shutdown all three Oconee units.
We have also reconsidered the initial DSI determination in the October 25, 1985 memorandum that resolution of this issue does not constitute e plant specific backfit. Contrary to that finding, it is our opinion that because the original licensing basis for Oconee did not include campliance with GDC 2 and 34 for the AFWS, this issue is a backfit. It appears clear that Generic Letter 81-14 itself is a staff approved generic backfit of these GOC. There-fore, because we see no means of resolving this issue without modifications at Oconee, we are proposing a plant specific backfit. We recommend this action despite the staff's previous approval of the SSF which omitted consideration of the single failure criterion (memorandum L. S. Rubenstein to Gus C. 1.ainas dated December 29, 1982 Enclosure 2) with only a single pump in order to achieve comparable post-earthquake safe shutdown capability to that provided at other PWRs in accordance with the criteria of Generic Letter 81-14 We further believe this action should be taken in view of the importance of both earthquakes and the auxiliary feedwater system to the risk of core melt.
We understand that before a backfit letter is transmitted to the licensee, the staff must develop a justification with the appropriate regulatory analysis (cost /benefitevaluation). We believt we can provide this information in a reasonable time frame with assistance from DSR0 (RRAB). However, before we proceed with this effort, we request your approval of our position. We are available to discuss this matter with you further if desired.
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Dennis Crutch iel , sist nt Director for Technical Sup rt Division of PWR Licensing-B
Enclosures:
As Stated cc w/ enclosure:
T. Schroeder J. Stol2
- 0. Parr G. Holahan F. Congel J. Wermiel H. Nicholaras J. T. Beard
ENCLOSURE 1 j
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UNITED $TATES NUCLEAR REGULATORY COMMISSION 2., J,i.: $' ! ' ,;
l j WASHINGTON, D. C. 20555 i Ob *Q,.:/ ' or it MB5 MEMORANDUM FOR: Gus C. Lainas, Assistant Director for Operating Reactors, Div- ici of Lin nsing FROM: L. S. Rubenstein, Assistant Director for Core and Plant Systems, Division of Systems Integration
SUBJECT:
SEISMIC QUALIFICATION OF THE AUXILIARY FEEDWATER SYSTEM AT THE OCONEE NUCLEAR PLANT, UNITS 1, 2 AND 3 - TAC N05.
43643, 43644, AND 43645 - SAFETY EVALUATION REPORT Enclosure 1 is a copy of our Safety Evaluation Report for Oconee, Units 1, 2 and 3 which was developed based on the licensee's response to the February 10, 1981 letter on Multiplant Action Item C-14 " Seismic Qualification of Auxiliary 1 feedwater Systems," and the TER, which was developed by Lawrence livermore National Laboratory (LLNL). The TER is an attachment to our Safety Evaluation Report aid is considered a part of our SER. The open items in our report involve the following areas:
- 1. Capability of the auxiliary feedwater system and/or safe shutdown facility to withstand a safe shutdow earthquake concurrent with a single active failure. This issue was previously addressed in the memorandum from L. Rubenstein to G. Lainas dated June 6, 1984 and in the memorandum from O. Parr to L. Rubenstein dated June 27, 1984.
- 2. Requirements for the isolation boundary between seismic and nonseismic portions of the AFWS.
- 3. Walkdown of the currently nonseismically qualified areas of the AFWS.
In our SSER for NUREG-0737, Item 11.E.1.1 for Oconee 1, 2 & 3 we stated that there was only one open item regarding the capability of Oconee to deliver AFW ficw following a seismic event or a tornado. Our evaluation of the seismic capebility (MPA C-14) of the Oconee AFWS is contained in Enclosure 1. In order to demonstrate the acceptability of the Oconee AFWS, which is not designed to withstand tornado generated missiles, we require acceptable responses to the attached questions contained in Enclosure 2.
Any modifications that are required to upgrade tne capability of Oconee to deliver AFW flow following a seismic event or a tornado are not considered to be "backfits" in accordance with Draft Manual Chapter 0514, *NRC Program for Management of Plant Specific Backfitting of Nuclear Power Plants." NUREG-0737 required a deterministic review of the AFWS using the acceptance criteria of SRP Section 10.4.9. In addition, Generic Letter 81-14 (MPA C-14) required
Contact:
P. Hearn X29461
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- 2 1985 cor'ormance with GDC 2 and 34 of Appendix A to 10 CFR 50. These are existico, fusiy-approved Commissior. requirements, not nev staff-imposed requirements. A copy of our SALP input is enclosed.
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L. S. Rulie stein, Assistant Director for Core and Plant Systems Division of Systems Integration
Enclosures:
As Stated i cc w/o enclosures: <
R. Bernero H. Thompson O. Parr J. Wilson J. Wermiel cc w/ enclosures:
J. Stolz R. Anand O. Thompson H. Nicholaras P. Hearn J. T. Beard
9 SALP INPUT Plant: Oconee
- 1. Management involvem mt.and Control in Assuring Quality: Not Applicable
- 2. Approach to Resolut1..n of Technical Issues from a Safety Standp Int:
Category 3 The licensee approach to resolving the issue of seis ic qualification of auxiliary feedwater systen was insufficient and showed little appreciation for the importance of the auxiliery feedwater system and the consequences of staring of systems between units.
- 3. Responsiveness to NRC Initiatives: Category 3 The licensee required an excessively long time to respond to our initiatives. The responses were also lacking important details.
- 4. Enforcement History: Not Applicable S. Reporting and Analysis of Reportable Events: Not Applicable
- 6. Staff (Including Management): Not Applicable
- 7. . Training and Qualification Effectiveness: Not Applicable i
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L SAFETY EVALUATION REPORT OCONEE UNITS 1, 2 AND 3 SEISMIC QUALIFICATION OF THE AUXILIARY FEEDWATER SYSTEM ntrodutfion l
Since the accident at Three Mile Island, attention has been focused on the ability of pressurized water reactors to provide reliable decay heat removal.-
While it is recognized that alternate methods may be available to remove decay heat following transients or accidents, heat removal via the steam generators is the first choice for accomplishing a safe shutdown of the plant. Therefore, there should be reasonable assurance that the auxiliary feedwater system (AFW) can withstand the postulated safe shutdown earthquake (SSE).
To address this concern, the NRC developed and initiated Multiplant Action C-14. " Seismic Qualification of Auxiliary feedwater Systems." The objective of this plan is to increase, to the extent practicable, the capability of those plants without seismically qualified AFW to withstand earthquakes up to the SSE level. This program was implemented with the issuance of HRC Generic Letter 81-14 dated February 10, 1961. Our review of the licensee's responses to this letter is the subject of this evaluation.
Evaluation The enclosed Technical Evaluation Report (TER) was prepared for us by our con-sultant, Lawrence Livermore National Laboratory, as part of our technical assistance contract program. The report provides their technical evaluation of the licensee's conformance to the requirements of Generic Letter 81-14. The A consultant's report indicates that the AFW may not continue to function during and following a seismic event as great as the safe shutdown earthquake. This conclusion is based upon cited weaknesses in the pumps, piping, valves, initia-tion and control, and structures / housing. The TER also indicates that the licensee did not conduct a walkdown of the AFW system and did not describe any alternate methods currently available to remove decay heat.
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Subsequent to the consultant's review, we requested the 1. ensee, in a letter dat- September 8, 1982, to review the c 7sultant's report and provide any corments relevant to our reaching a safety conclusion. The licensee's response, dated October 13, 1982, emphasized their belief that the AFW does have substantial seismic capability in that it would remain functional following'an operating basis earthquake (i.e, half the level of the SSE)'.
The response also requested additional consideration of a fully seismically qualified dedicated shutdown facility, and provided specific comments and information. We have reviewed this supplemental information provided by the licensee, our consultant's technical evaluation, and have performed our own review of the licensee's responses to Generic Letter 81-14 and our request for additional information. Our summary findings are described below.
Pumps and intors The turbine-driven AFW pump could fail during a seismic event due to the loss of one of its support systems. There is no retrievable documentation on the seismic capability of the turbine oil system, although the turbine, as a whole, was certified by its manufacturer. The other trains of the AFW include.
two full capacity seismic Category I electric motor-driven pumps per reactor.
Therefore, the potential seismic failure of the turbine-driven AFW train is acceptable on the basis of sufficient unaffected redundancy. (Thatis,the two motor-driven pumps will be operable). The housing of the pumps in the turbine building is discussed later.
Piping The piping for the AFW systems is seismically qualified to the SSE level out through the first isolation valves, which are normally closed. Piping beyond these boundary points is not currently seismically qualified. The licensee indicates that this situation is consistent with other safety-related systems at the Oconee station.
3 Gen < ic Letter 81-14 requests licensees t( consider 'he t AFW systems as including piping up to and including the second valve which is not ally closed, or capable of automatic closure when the isolation function is required. This system boundary definition is intended to assure that the safety function of the AFW will not be lost during a seismic event, assuming that the seismic event causes the failure of the nonqualified piping concurrent with a single failure in the isolation valve.
Thelicenseehasidentifiedthelowpressureservicewatersystem(LPSWS)and portions of tha AFW system where the boundary between the seismic and nonseismic portions are separated by a single manual isolation valve. The LPSWS provides cooling to the two AFW motor-driven pumps.
The licentee has agreed to upgrade or replace any of the AFW isolation valves that are not qualified to remain functional after a safe shutdown earthquake.
In their analysis, the licensee has assumed that a single active failure will not change safely positioned manual isolation valves to the unsafe' position.
This assumption is correct only if the following conditions exist: 1) the valve position 15 inspected every 30 days or after . valve position changes or repairs, and 2) the circuit breaker to any electrical controls for the valve operator is opened and the breaker position is inspected every 30 days or after position changes or repairs. The licensee refuses to open the valve motor operator breakers and proposes to inspect manual valve positions every 90 days or after repairs. If the valve motor operator breakers are not opened and the breakers
- and valves' are not inspected every 30 days or after (valve or breaker) position changes or repairs, we conclude that an SSE concurrence with a single active failure in these isolation valves would result in the loss of all of the AFW system. The loss of all of the AFWS would result from the draining of the AFWS through a break in the nonseismic portion of the AFWS.
Because of the serious consequences that would result from a total loss of the AFWS, it is our position that the licensee assure that a single active failure will not change a safely positioned manual isolation valve to the unsafe position. The licensee can satisfy this position by meeting the following conditlens:
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- 1. Inspect the manual vr ve p' ions every 30 days or aftar valve position chang (s or repairs, and
- 2. Open the circuit breaker to any electrical controls for the valve and inspect the breaker position every 30 days or after position changes or repairs.
The licensee contends that the LPSWS is capable of functioning af ter a safe shutdown earthquake but had made this conclusion without a detailed analysis, in order for us to complete our evaluation of the LPSWS we will require the licensee to submit a detailed justification for concluding that the LPSWS will function after a safe shutdown earthquake. ,
The licensee is analyzing the effects of a safe shutdown earthquake on plant heating lines, failure of which may affect the functional capability of the AFWS. The licensee has comitted to modify these heating lines as required.
In order to complete our review we need to know the results of the licensee's analysis of the plant heating lines, including a description of any modifications to these lines.
Valves and Actuators j
The following are the only valves in the AFW that are not qualified for the I
SSE. !
- 2. The air-operated valves are not fully qualified. ;
The licensee has indicated that the areas lacking qualification have no effect on the operability of the AFW. All the oil valves that support the AFW are '
related to the turbine-driven pump. These valves are acceptable on the basis that the plant can be placed in the cold shutdown condition without the turbine-driven pump.
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With regard t( the concern for potential failure of the .ir-operated valves, only two valve in the AFW system per unit " t change position to establish and/or control flow to the steam generators. These valves are air-operated, are normally closed, and fail to the open position. Documentation on the seismic qualification of these valves is not available. In order to provide assurarce that these valves will be capable of operating following an SSE, the licensee plans to qualify these valves either by analysis or by replacement, as required. Based on the licensee's commitments, we conclude that the !
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auxiliary feedwater system valves and actuator are capable of functioning
! after a safe shutdown earthquake.
The air operated valve #C-176 isolates the line that connects the suction'line of the AFW pumps to the main condenser hotwell. Within 30 to 40 minutes of a I
loss of the air supply, valve #C-176 will open and begin to drain the upper surge tank (the primary water source for the AFWS) into the main condenser hotwell. This will result in starving the flow to the AFWS approximately 18 minutes after valve #C-176 opens. The opera'ttng procedures instruct the operator upon the loss of the air supply to close valve #C-176 and align the AFWS to the alternate water supply which is the condenser hotwell. The l j operator has between 48 and 58 minutes to perform the operations that prevent starving the flow to the AFWS. This is well within the operator action time requirements (30 minutes); therefore, we conclude that the licensee's method '
for preventing the starving of flow to the AFWS due to the loss of air supply to valve eC-176 is acceptable.
Power Supplies f
Electric power to some of the motor-operated valves and pneumatic sources for For the MOVs, the licensee f air-operated valves are not seismically qualified.
stated that electric power is not essential since the MOVs fail as-is and are not required to change position to establish flow. While we agree that
' establishing AFW flow is acceptably ir. dependent of electric power, we remain concerned regarding control of AFW flow. The applicant has assured us that the motor-operated valves are not used for AFW flow control; therefore, For l electric power is not needed for them to perform their safety function.
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the air-operated valves, which includes the norral flow control-valves (FCVs) for the AFW, the licet ?e has provided an automatic bottled nitrogen system which can serve as an alternate to the air source. The licensee has comitted.
to assure that the automatic bottled nitrogen system, including power to the solenoid valves, will withstand a safe shutdown earthquake.
Based on the licensee's commitments, we conclude that the AFWS power supplies are capable of functioning after a safe shutdown earthquake.
Initiation and Control The control to the motor-operated valves other than those in the . auto-initia-tion and auto-control of the AFW system is not seismically qualified. This includes the control to the branch line isolation valves nff the main steam header and the electric motor-operated valves in the AFW suction and discharge line which are normally aligned for AFW operation but not normally required to operate. Huwever, the licensee stated that no actuation is required of the motor-operated valves for the AFW flow and the valves will fail as-is upon loss of power. In order for this design to be acceptable, the licensee must comit to open the breakers to these valve operators and verify that the breakers and valves are open every 30 days or after position changes or repairs.
Structures The turbine building which houses portions of the AFW system is seismic Class II, The licensee has reanalyzed the turbine building and determined that the structure will survive the safe shutdown earthquake. The licensee's reanalysis of the turbine building is presently being evaluated by the staff. ;
Standby Shutdown Facility The standby shutdown facility (SSF) system has been constructed to provide a dedicated separate train of auxiliary feedwater in the event the AFWS is simultaneously incapacitated on all three units by a safe shutdown earthquake.
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Thc SSF system is designed tr ithstand the SSE. Structures suppor+',g or housing the SSE system componei s include the reactor building and axiliary building and are seismic Class . The licensee previded a description of the methodologies and acceptance criteria used for seismic qualification of the SSF system, referring to applicable sections of the FSAR and licensee's letters of March 28, 1980; February 16, 1981; March 31, 1981; and April 13, 1981.
Regarding the AFW system boundary, all connected branch piping and crossover connections among the three units are seismically qualified only through the first valve. We conclude that the AFW system boundary does not fully meet the requirements defined in the Generic letter.
Regarding the system boundary, some small piping vents and drains, capped lines, tank vents, and a recirculation line from the diesel fuel oil storage tank cither have only one normally closed valve or are seismically designed through the first valve. We conclude that the SSF system boundary does not conform to the definition of boundary specified in the Generic Letter.
Furthermore, we require that this deviation be evaluated and corrected in order to assure the required safety function of the SSF system.
q Our consultant has made the following conclusions regarding the SSF.
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- 1. The licensee did not perform a walkdown of the currently nonseismically qualified areas of the AFW system because the SSF system is designed to withstand the SSE and to serve as the alternate decay heat removal system.
We do not fully concur with our consultant's conclusion that the SSF is a sub-stitute for the AFWS. In order for the SSF to be considered a substitute i
for the AFW, it would have to be capable of withstanding an SSE concurrent with a single active f&ilure. The SSF contains only one auxiliary feedwater l l
pump; therefore, if the SSE disables the AFWS for all three units and a single l l
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1 active fa"ure ' sable- the SSF AFW pump, all feedi ter to the three units i 9
would be lost.
The licensee not only overlooked the single active failure disabling the SSF AFW pump, but the use of one SSF AFW pump to provide feedwater to all three i
Oconee units violates' General Design Criterion Number 5, which states:
" Structures, systems and components important to safety shall not be shared among nuclear power units unless it can be shown that such L
sharing will not significantly impair their ability to perform their safety functions, including in the event of an accident in one unit, ar orderly shutdown and cooldown of the remaining units."
In the event of an SSE-induced accident, the licensee cannot show that an orderly shutdown of the remaining two units can be perfonned because t'he SSE can cause an accident in all three units simultaneously. This situation arises when an SSE ruptures the circulating weter line and the resulting internal flood disables all the AFW pumps in the three units. The result is an accident in all three units, namely, the loss of the ultimate heat sink..
The safety significance of connecting the three units to one SSE AFW pump is any perturbation in one unit will cause perturbations in the other two units.
Considering tnat Oconee is a Babcock and Wilcox reactor, thereby extremely sensitive to perturbations in AFW flow, small perturbations in AFW flow in one General unit could cause severe transients or accidents in the other two units.
Design Criterion Number 5 was enacted to avoid these very plant interactions.
Because simultaneously putting the three Oconee units on one SSF AFW pump violates General Design Criterion Number 5, it is our position that the licensee provide procedures that restricts the use of the SSF AFW system to one unit for each accident or event. The licensee can satisfy this conditiun by meeting the following conditions:
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- 1. Qualifying the AFW system to remain functiont ' after a safe shutdown earthquake, and
- 2. Providing watertight doors, barriers and enclosures between the circulating water lines and the AFW system in order to prevent a seismically-induced l circulating water failure from disabling the AFW system for the three )
Oconee units.
1 In summary, our evaluation concludes that the licensee's AFW system does not ]
j possess an overall seismic capability for withstanding an SSE. This oversight can easily be corrected by minor plant modifications and rewriting some opera-ting procedures. Because of the importance of the AFW system, we believe the qualification of the AFW system to the Seismic Category I classification should ]
be a high priority for the licensee, i
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. August 27, l'82 l TECHNICAL EVALUATION REPDR'.
E, 'EE N'JCLEAR ST ATI?' UNITS ( 2, Ag j J
SEISMIC 6 _IF.' ATION Dr & ILI A:Y FEEDWATER SYSTEM
- 1. IC RDDUCTION Sin:e the accident at Three Mile Islano, considerable attention has been f t:use: or. the capability of nuclear power plants to reliably remove de:ay heat. Tne M: has recently undertaken >taltiplant Action Plan C-14 "Seistic Leslification of Ari Systems" [Ref.1), which is the suoject of this evslua-
- ion.
To agement the first phase of Action Plan C-14, the NRC issue: Generic Letter NO. El-14 "Seis?.it Qualification of ArW Systems" [Re'. 2), dated Tet:uary 3C,19E1, tc a21 operating PWR licensees. This lette recuested ea:h licensee (1) to condu:t a walk-down of non-seismically cualifie: pc:tions cf the AFW system and identify Deficiencies amenable to simple actions to improve seisaic resistance, and (2) te provide design information rega-ding the seis-ni: capao!!!ty cf tne AFW system to facilitate PRC backfit oe:!sions.
The licensee of 0:onee No:les: Station responded with a letter dated January 28,1962 [Ref. 33. The licensee's response was found not to be com-plete and a Request for Additional Information was issued by the N;C, cate:'
Ap:11 6, 19E2 [Ref. 43 The licensee provided a supplemental response in a letter dated May 25,1982 [Ref. 5).
This report provides a technical evaluation of the information rovided in the licensee's responses to the Generic Letter, and includes a recommendation regarding the need for additional analysis and/or upgrading inodification of this plant's AFW system.
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Informati- provided in licensee's responses incluoed:
o Specifict:lon of the overall seismic capability of the AFW system.
o Identification of AFW system components that are currently non-seismically cualified fo: SSE.
o Discussion of levels of seismic capability of non-seismically 1 cualifice components, o Spe:ifi:stion of ove:all seismic caoability of the Standby Shutoo=n Fa:ility (SSF) system which will serve as an alternate ce ay heat ]
removal system.
o Des::!ptio . of methodolo;1es and acceptance criteria for the seistic q design of the SSF system, which is ceterminec to be seismically I cualifier fer the SSE level by the licensee.
o Description of the ATW and SS system Doundary.
o Status of compliance with seismi: related N C BJ11etins and Inicrma- ,
t!Cn Notices.
o Ao:itionally, schematic sketches of the ArW and SSF r fstems.
o A::ditionally, identification of areas of mo:ification of the AN system that will be perfcrmed under the SSF proje:t.
o Ac:!tionally, description of method:logies and acceptan:e criteria for seismically qualified components of the AFW system.
We have reviewed the licensee's responses, and a point-by-point evaluation of licensee's responses against Generic Letter's requirements is provided !
1 btlow.
(1) Seismic Capability of AFW System ,,
Except for those items identified in the following, the AFW system has been designed, constructed and maintained to withstand an SSE I utilizing methods and acceptance criteria consistent with that applicable ,
to other safety-related systems in the plant. Presently, those items l identified by the licensee as not being fully seismically Qualified are evaluated below:
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7 c F.m:s/M-tcis - Portions of the turbine-driven pump til system and oil i corling system, including the oil pt :s and water cooling pumps do n;t have retries.cle seismic occumen ation. However, we judge by l 1
experience that the ;, umps / motors possess e less : nan CEE level of !
seismic csostity. )j o Pipino - The portion of all connected branch piping beyond the first valve is currently non-seismically cualified. We believe that the Arm system piping is likely to possess an DBE level cf seismic CEDE *ity.
c valves /A::usicrs - (s) Oil valves in the support system. However, j the licensee in icated that credit for seismic design is not neces-SEry becsuse they are equipped with handwheels for manual oper6-tions. (c) PnBJmatic Control valves and their backup nitrog!D brttles. Ho ever, the licensee inoiCat-:$ that these vElves will fell open upcn Icss of gas pressure or they can be bypassed by sligning the Arm flo. throug% the mein feedwater startup line into the normEl or Arm steEm generatC: nor21es on either steam generator.
(c) Certain valves do not have retrievable seismic documentation. ]
The licensee stated, however, that such valves were built to at least ]
the ANSI B 31.1.0 criteria and were modeled into the stress analyses
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as eovivalent pieces of pipe for structural purposes. Based on the j
. Eteve information, we believe that the valves / actuators are likely to possess an DBE level of seismic capacity. '
o Power Suo:11es - Power to the electric motor-operEted valves and pumps, except for the motor-driven AFW pmps anc the lower pressure service water pumps, is currently non-seismically qualified. How-ever, the licensee stated that seismic design credit is not necessary for the power to the electric motor-operated valver because these valves can be manually operated with handwheels. We judge that the power supplies possess a less than CBE level of seismic c%pacity.
o Water Source (s) - None o Initiation / Control Systems - The control to the motor-operated valves l other than those in the auto-initiation and auto-control of the AFW system is not seismically qualdfied. This includes the control to the branch line isolation valves off the main steam heacer and the electric motor-operated valves in the AFW suction and discharge lines which are normally aligned for AFW operation but not normally recuired to operate. However, the licensee stated that no actuation I
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i; recuired of the motor-operated valves for the a flow and the j valves will fail as-is upon loss of power. We therefore iudge that I d
the initiation / control systets possess the capacity te wi, stant an j SSE.
f o Structurer - The turbine builcing is seismic Class 11. We therefore judge that the structures supporting or hcusing th. ATW system co9ponents are capable of withstanding an DBE.
Based on our evaluation, those areas of the AFW system judged not tc
. possess an SSE seistic capatility are identified below: j o Punos/ Motors Less than DBE I e F1:!n; OBE o Valves /Actusters DBE o Fo er Supolies Less than DBE o ha:e; Source (s) None o 2nitiation/Contic1 Svstens None o Structures DBE In summary, our evaluation indicated that the licensee's AFW system coes not Ocssess an overall seismic capability that can withstat.d an SSE.
Be:asse the primary wate source is seismically cualifiec for the SSE, a switchover to a seismically Qualified secondary water source is not involved.
The Standoy Shutdown Facility (SSF) system, being constructed to provide a oedicated separate train of auxiliary feedwater, will provide an alternate decay heat removal system when it becomes operational. -
No procedure is available at this time to switch from the AFW syst'em to the SSF systes. Such procedure will be developed on a schedule commensurate with the SSF system startup. The licensee did not indicate the completion date of the new SSF system. I J
l The SST system is designed to withstand the SSE. Structures ;
supporting or housing the SSF system canponents include the reactor building and auxiliary building and are seismic Class I. The licensee's 3 provided a description of the methodologies and acceptance criteria used p
fer seismic ouall'ication of the SSF system, referring to applicable sections of t*e FSAR and licensee's letters of March 28, 1980; Fe:.uary 16, 1981; March 31, 1911; and April 13, 1981.
Regarding the TW system bnundsry, all connected bran 0h piping and crossover connecti:ns among tht thIde Units are seisticElly Qualified only through the first valve. e judge that the AFW system bou':ary does not fully meet tne reasirements defined in the Generic Letter.
Regarcin; the SSF system boundary, some small piping vents and trains, capper lines, tank vents, and a recirculation line from the diesel fuel oil stcrage tank either have only one normally closed valve or are seismically cesigned only through the first valve. We judge that the SSF sys~em DDuncary does not conform to the definjtion of boJhda!y specified in the Gene !C Letter. Since the existing AFW system is not fully seismically cualified, we feel thzt this deviation needs to be evaluated an~/Cr C0rreCte: in orde! te assur$ the 780uiIed safety function of the S5rsy3:em, Tne licensee stated that both the AFW anc SS~ systems we:e included within the scope of the seismic related NRC Balletins 79-02, 79-04, 75-07, 75-14, 80-11, and IE Information Notice 80-21.
(2) Walk-Down of Non-Seismically Dualified Portions of ArW System The licensee stated that no walk-down was performed for the non-seismically qualified items of the AFW system due to reliance on the SSF system though the walk-down is requested by GL 81-14 We feel that a walk-down is required if the new SSF system does not become operational within a reasonable period of time.
(3) Additional Information The licensee provided a schematic sketch of the AFW and SSF systems including the water source (s), heat sink, suction and discharge piping, major mechanical equipment, and structures support.ng and housing the AFW and SSF system items.
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. ;c:itionally, licensee's responses provided a description of the methodologies and acceptance criteria that were used in the design of the seismically cualified portions of the AFW system, by referring to the applicaDie se:tions in the FSAR. ,
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i The licenset identified the areas of the ArW system where modifice-tion /uograde will be performed for the tie-in between the SSF and AFW J
systees. Because the construction of the SSF system is underway, the !
licensee stated that no additional modification to the AFW systen is ne:essary ove to reliance uoon the SSF system.
- 3. CDNO_UE 0NS l T .e i-ftrmation contained jn licensee's responses is complete. The licensee cic not perferm walk-down of the currently non-seismically cualified areas of the AFW system be:ause the SSF system, being unoer construction,' is designed to withs*.and the SSE and to serve as the ,a.1 ternate decay heat removal system. The switchover procedure from the AFW to the SSF system will be established commensurate with the startup operation of the SSF systeh. Both tre Ark ard SSr system boundaries co not fully meet the definition spe:Afded in CL El-14 SEsec upon the submitted information, we conclude tha'. the A't system coes not presently possess the seismic capatility to withstand an SSE. The ability of the SSF system to perform the required safety function following the occurence of an SSE is also in question because the SSF syster.. boundary does not fully conform to the boundary definition spe:ified in GL 81-14. In conclusion, we recommend that the NRC considers requiring the licensee (a) to submit the estimated completion date of the SSF system and perform e walk-down of the existing AFW system if it is determined that the SSF systgm will not become operational within a reasonable period of time and (b) to evaluate and/or correct the deviation of the SSF system boundary in order to essure the required safety related function. '
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REFERENCES
' 1. D. C. Eisenhut, U.S. Nuclear Regulatory Commission, memorandum to H. R.
Denton, "Multiplar. Action Plan C-14: Seismic Qualifica' .on of Auxiliary Feedvater Syste 3," February 20,19! . .
- 2. U.S. Nuclear Regulatory Commission, Generi: Letter No. 81-14 to all opere- "
- ; pressurized wate
- reactor licensees " Seismic Qualification of Auxil-Isry Feedaater Systems," February 10, 1981.
- 3. h. O. Parke;, Jr., Duke Power Company, letter to H. R. Denton of U.S.
Nuclear Regulatory Commission, January 28, 1982.
- 4. J. F. 51c12, U.S. NJ: lear Regulatory Commission, letter to W. O. Parker, Jr., of Duke Po=sr Company, " Request for Additions 1 Information on Seistic I Qualification of the Auxiliary Feeo< ster System, Dconee Nuclear Station 1 Units 1, 2, and 3, April 6,1982.
- 5. W. O. FE:ker, Jr., Duke Power Company, letter to H. R. Denton, U.S.
Nuclear Regulate:y Commission, May 25, 1982.
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ENCLOSURE II REQUEST FOR ADDITIONAL INFORMATION OCONEE NUCLEAR STATION f,UXILIARY SYSTEMS BRANCH I
- 1. Provide a probabilistic risk assessment (PRA) which demo ttrates that the probability of significant damage to the emergency feedwater system, follom ng a missile strike resulting from tornadoes and other high winds andassuminglos9ofoffsitepower,islessthanoregualtoamedian value of 1 x 10~ per year or a mean value of 1 x 10' per year. Also provide a description of the methodology, modeling, assumptions, and error bounds used in your PRA.
Your PRA should use the probability of windspeeds that is provided in Figure 1. Significant damage is defined as damage that would prevent meeting the design basis safety function.
- 2. You stated that in the event of a tornado induced loss of both the main and emergency feedwater system, the steam generator inventory would be expected to be boiled off within a few minutes. Ample time wou h' he available for opening the manual dump valves on at least one steam generator to maintain a low back pressure for the auxiliary service water pump. Blowdown of the steam generators would not be necessary.
Provide the results of an analysis which demonstrates that adequate decay heat removal can be continuously maintained through the use of the existing auxiliary service water system, and that such a cooldown method will not result in an accidental overpressurization of the auxiliary service water system c- the excessive loss of reactor coolant. As a minfmum, the followir.g points should be addressed:
- a. Since the effectiveness of water injection into the steam generator (SG) is of some concern because of the low heat capacity of the ASWS, discuss the SG pressure that must be attained in order to provide sufficient ASWS flow into the SG to ensure adequate decay heat removal. Further, prcvide the dump valve capacity at rate ASWS pres-sure, and the time period assumed for operator action of the manual dump valves.
- b. Provide a discussion and analytical results (plots if appropriate) of the transient following a reactor trip utilizing the ASWS to remove decay heat. The discussion should include reactor coolant system pres-sure and temperature, SG pressure, SG water level / flow rate, and decay heat removal versus time. Provide the time at which the secondary steam dump capacity will match decay heat load following a reactor trip provided that the steam generators are at the ASWS operating pressure. Following the reactor trip during the time for which decay heat is greater than the removal capability of the steam dumps at ASWS pressure, provide the mass loss through the reactor system safety valves.
- c. Discuss the effects of cold shocking the steam generator as a result of injecting cold water into a relatively dry steam generator.
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- EN, CLOSURE 2 UNITED STATES
,' ~; NUCLE-R REGULATORY COMt.*lSE ON e'. v j WASHINGTON. C,. C 20155
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~9 N D m e: Los. 50-269 50-270 50-287 MSMCRA CUN F0F.: Gus C. Lainas, Assistant Director for Operating Reactors, Division of Licensing, NRR FROM: L. S. Rubenstein, Assistant Director for Core and Plant Systems, Division of Systems Integration, NRR
SUBJECT:
SAFETY EVALUATION REPORT FOR OCONEE NUCLEAR STATION, STANDBY SHUTDOWN FACILITY AND SECTICNS III.G.3 A"O III.L OF APPENDIX R 7010 CFR 50 Enclosed input regarding is the theAuxiliary Systems Oconee Standby Branch Facility Shutdown (ASB) SSF).
Safety (Evaluation This facility wasReport (SER provided by the licensee to resolve concer,ns regarding safe shutdown of the Oconee Nuclear Station following a fire, internal flooding of the turbir,e building, and security incidents. This SER only addresses the concerns of safe shutcown in the event of a fire and internal flooding. The aspect of security will be handled by others.
Regarding the safe shutdown capability in the event of fire, we conclude that the Oconee Nuclear Station inclucing the SSF, is in conformance with the re:;uirements of Sections III.G.3 and III.L of Appendix R with the exceotien of providing source range flux and steam generator prMsure indication at the 53F.
It is our position that such instru~entation is reevired. The enclosed SER also resolves the open items in our SER dateo August 22, 1978. ,
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l Regarding the safe shutdown capability in the event of internal flooding of the turbine building, the licensee has demonstrated that the three Oconee units can l l
be safely shut downutilizing the SSF. l In agreement with the DL project managar Phil Wagner) ASB has incorporated the l SER inputs of the DSI branches involvec. Inouts from CE will be coordinated by l l
DL.
Pending receipt of a commitment by the licensee to provide the instrumentation ,
noted above, we conclude that our review of the Oconee Standby Shutdown Facility and the compliance of the Oconee Nuclear Station with the requirements of Sections III.G.3 and III.L of Appendix R is c,omplete.
.% :Tl W A.c L. S.'R benstein, Assistant Director for Core and Plant Systems cu S g2 O ..
. ~.l O G Q H .7f~ i Division of Systems Integration Enclosure and cc:
See next page
Contact:
l T. Chan, X29460 L ___
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A: Statec c: ></ e nci c ". u re :
P. . Mattson R. Cacra D. Eisenhut O. Farr V. Panciera R. Lobel V. Senaroya T. War. bach J. Stol:
P. Wagner-c
?i . Connor
't. Fioravante T. Chan X. Luckas(BNL)
F. Rose T. Dunning R. Karsch M. Srinivasan J. E. Knignt S. Rhow I
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'SAyETY EVALUATION REPORT .
1 OCONEE NUCLEAR STATION !
STANL3Y SHUTtJWN FACILITY AUXIL'**" $YSTEP.S DKANCH
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By letter dated February 1,1976 t he Licensee proposed a-safe Such a system I
shutdeun system for the Oconee Nuclear Station. ,,
wculd augment existing plant ca abilities relative t'o miti- -
such as fires, security-ir.ci-gating postulated occurrences dents and turbine building flooding. Ad:itional i n f o r,ma t i on describing the conceptual design of the safe shutdown system was received by letter dr:ed June 19, 1978; subsecuent staff approval of the conceptual design was transmitted to the, Licensee December 29, 1978. In accordance with the conceptual design evaluation, the licensee provided a final design pro-pos a l f or the system, the standby shutdown facility (SSF), in _
At the time of the March 28, 1980 a March 28, 1980 submittal. t suhnittst, Appendix R was not effective. .
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On February 19, 1981, the fire protection rule for nuclear power plants, Appendix R to 10 CFR 50 became effective.
This rule recuired att licensees of plantslicensed prior to .
January 1, 1979, to submit by March 19, 1981: (1) plans and schedules f or meeting the applicable ree.uirements of . Appendix R, a
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(2) a design description of any morificat ons proposed to prov.de alternative safe sht c;. n capability pursuant to c arag-eph III.G.3 of Appendix R, and (3) exemptson recuests fer which the telling provisions of Section 50.48(c)(6) was to be invoked.Section III.G of Appendix R is a retrofit item to all pre-1979 plants regardless of previouf SER cositions and resolutions. Subsequently, the l i c e'n s e e proviced submittals reDarding the use of the SSF to meet Appendix R requirements for the Oconee Nuclear Station.
, It should be noted that this SER only addresses the ' concerns of safe shutdown in the event of fires and turbine building flooding. Saf e shutdown in the event of a security incident will be handled by others.
The licensee has addressed the Oconee Nuclear Station's post-fire shutdown capability in six letters dated January 25, February 1, and June 19, 1978, March 28, 1980, and March '18 and April 30, 1981. Additional inf ormation was provided in letters dated January 25, and September 20, 1982. These submi:ttals discuss the various means used to achieve and main-tain saf e shutdown conditions, determine whether saf e shutdown could be achieved without equ' ament or cabling in any one fire area, and identify any modification required due to unacceptable interactions caused by a fireI
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- 2. (SJ, SYSTEMS DE5_CRIPTION 1
2.1 Mechanical Systems-2.1.1 General Descriptieb 1
The standby shutdown facility (SSF) is a " bunkered" facility which hous,es the systems and components neces'sar'y to provide an alternate and independent means to achieve and maintain a het shutdown condition..for one or more of the three Oconee l units. The SSF serves to resolve the safe'shutdovn I requirement for fire protection, physical security and turbine building flooding. The SSF has the capability of maintaining hot shutdown conditions in all three units for approximately three days following a loss of normat AL power.
The licensee concluded that the most likely reason for flooding of the turbine building would be from a condenser circulating water pipe break resulting from a seismic event. The licensee therefore decided that the SSF would be a seismic Category I structure and further that it 'o e destgned to withstand the effects of tornadoes. The missile spectrum upon which their analysis is based, is in confor-mance with the guidelines of the Standard Revis. Plan (SRP) )
Section 3.5.1.4, Revision 1, for a tornado Zone i site. The i
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- rade levet entrance elevatier. of the.S$F is 797.0 est.
This eles. : ion is below Keowee fu'l pond elevation of S00 s .el' 'as tr.e axir.u- 'eir elevation of 505.. -: '
.e , in the event of flooding due to a break of the non seismic con-l denser circulating water (CCW) system /pioing locat ed in the l
turbine building, the maximum expected water level within
- the site boundary is EL 796.5. Sincethemaxi[um exoected ater level is below the elevation of the grade leve,l entrance to the SSf, the structure will not be flooded by such an incident. In addition, the structure wiLL be waterproofed to prevent infiltration of normal ground water. Thus the structure meets the requirements of GDC 2, and the guidelines of Regula t o ry Gui de 1.102 wi t h respe ct to protection against flooding. Since the use of .he SSF may be recuired folle:ing a tornadic event to meet II.E.1.1 requirements, a' separate review of the structures and systems to withstand the effects -
1' of tornado missiles will be performed.
2.1 2 .neactor tenlant_(EC) Met ae S *a/
The S$r RC makeup system is designed to supply makeup to the reactor coolant system (RCS) in the event that normal systems are unavailable. The capacity of the system is sized to account for normal system leakage, and shrinkage which results from going from a hot power operating condi-tion to hot shutdown.
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~5-ine criter) component of the SSF RC makeup . <stec is the 26 gpm ESF RC pump. One pucp is required for each of the three ur.its; each pump is located in its respective reactor o
building. The capacity is sufficient to maintain RC inven-e tory during the transition from power operations to hot s h.;t d o w n. The makeup source is from the spent fuel pool, thus ensuring a supply of berated. water. Letdown, if required, is returned to the spent fuel pool. The letdown valve is powered from the SSF power system and is controlled from the SSF ohly. Capability for 'ower p and control of one bank of pressurizer' heaters atlow control of the steam bubble in the pressurizer. Overpressurization protection is provided by existing relief valves. The system is designed to seismic Category I and Quality Group B require-ments. Failure of the SSF RC nakeup components will not i
affect the operation of the normal "in plant" components.
The S$F RC makeup system is operated and/or tested only from the SSF.
2.1.3 StF Anv4f(nry te-v4ce Water System (SSF ASVS)
The SSF ASWS is a high head, high volume system designed to provide sufficient secondary side inventory for adequate decay, heat removal during a less of normat AC power (n conjunction with the toss nf the normal and eme'rgency feed-water systems. ,
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The SSF ASWS pump is the major compone:.t of the system, and is housed in the SSF. The one motor driver ump which is powered from the SSF power system is designed ,to provide approximat ely 750 gpm at futt system pressure to each of the three units for approximately three days. .The water contained in the embedded condenser circulating water (CCW) !
piping serves as the water supply. The embedded portion of the CCW piping is designed to withstand the effect.s of a seismic event!. The SSF ASWS is designed to seismic Category I and Quality Group B'and C requirements. Failure of the SSF ASWS components wilL not affect the operation of the normal "in plant" c o,m p o n e n t s . The SSF ASWS'is operated and/or tested only from the SSF.
enmum .
2.1 4 _sst snev4e. Water System, The SSF service water system is comprised of the HVAC ser-vice water system and the diesel engine service watet system.
The HVAC service water system, which operates continuously, contains two 1001' capacity pumps and supplies cooling water to the HVAC condensers. Only one pump operates at a given t'i m e , the other pump serves as a backvo.
The diesel engine service water system, which operates only when the diesel is operating, contains one pump and provides service water to the diesel engine jacket water heat exchan-gers.
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4 Al' three pc :s take their suction from the embec.ed CCW oicing and return the flow to the CCW piping after passing through their respective system. The piping and components the SSF service water system are designed to withstand cf seismic event. All pumps are powered the effects of a from the SSF power system..
2 . '. 5 _*** " "> r 'v"*- I The SSF HVAC system is comprised of a ventilation system '.
and an air conditioning system. The HVAC system provi' des !
filt ered and conditioned ventilation for the'SSF structure, and maintains the environmental conditions within the limits set for p e r s o n n ot..o c c up a n c y and e:uipment operability.
The combustion air and diesel engine exhaust s,y s t e m s are independent of the ventilation systems. The HVAC system consists of three ventilation fans, two air conditi6ning refrigerator, units and associated ductwork. All c e r.p o n e n t s and ductwork which service ar. i which contain ecuipment n'eeded for safe shutdown are ctsigned to withstand the effects of a seismic event. All components are powered from the 1
$$F power system.
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-S-2 . '. . d L51_ S.v".A Svt*em The SSF sump system provides a means of collecting and dis--
charging drainage from equipment and floor drains.. 7,w o sump pumps, both of which are powered from the SSF power system, discharge the effluent to the yard drainage system.
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The discharge line will be designed 30 prevent back flew into the sump.from outside sources. Level' switches L.o c a t e d in the sump are used to automatically start and stop the pumps. The sump system io a nonseismic system as its failure will not adversely' affect equipment needed for safe shutdown.
2d.7 .LS.Efotable yat.e Svst.m The potable water system provides potable water 1,or sani-tation and potable services. Water is provided from a 200 gallon storage tank located in the SST. The tank is fed from the plant potable water system, and tank level is float' controlled. Water to t he 6.)r t able b a t t e ry test facility is supplied directly from the plant petable water stem rather than from the s.torage tank. The sys tem is . designed to nonseismic requirements since its failure vill not adversely af f ect equipment needed for safe s h ut down.
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. 2.1.! : instrumentation and Con,t21; t
The SSF control panel prnvides instrumentation and controls l
' needed to assure' safe plant operations and shutdown condi-Monitoring capability is pro-tions for the three units.
vided for plant parameters such as primary coolant tempera-ture and pressure, pressurizer pressure and Level, incore i
thermocouple, and steam generator level, and diagnostic (
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capabilities ,for mechanical and electrical component per-However, steam generator pressure and neutron "
formance.
1 flux indication, and steam generator pressure controls have not been provided in the SSF. AlL electrical equip-ment necessary for hot shutdown is designed to withstand Electrical power is pro-the effects of a seismic event.
vided by the SSF power system.
2.2 SSF Power Eve +*a 2.2.1 General Description The S$F power system is comprised of independent emergency sources of AC and DC electrical power and their associated electrical distr .f ution sys t ems, and various support systems.
It would be operated only in the event installed normat standby systems are inoperable. Manual operator action is recoired to actuate the system.
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l-t The SSF ocwer sy s t em include: onsite 4160VAC, 60C"AC, 120VAC are 125VDC power. This syste- suc -ties power necessary for
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. c' tne reatt:- in the event of !:ss c' cower from all other power systems. It consists of switch-g <e a r , a lead cent er, a motor control center, panelboards, battery, battery chargers, an inverter, a diesel-eleetrical generator unit, relays, control devices, and interconnecting
- s Le suoplying the appropriate loads.
The inverter supplied 120VAC power system s upp ly i ng the security system circuitt in conjunction with the 125VDC instrumentation and control power system supplies continuous control power to all loads that are recuired for a hot shutdown of the reactor.
2.2.2 AC Power Interfacingm The 4160 volt SSF power system is provided for backup ser-vi c e only and is nor ally de-energi:ed with alt breakers en the bus in the open position. Upon lots of the normal p:wer system, the 4160 VAC SSF power system will provide power to the necessary loads to saf ely s hut.down the unit by an onsite diesel-electric generator (see Section 2.2.3 for further discussion) which is independent of the normal power distribution system. All of the loads required for hot shutdown of the reactor are supplied power during loss of the normal distribut'.on system from the 4160 VAG SSF power system, either directly (for the high head auxiliary se rvi ce water pump) or through t ransf ormer (s) if at a lower 7
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a The 660VAC SSF load center is normally de-energi.ed. Power to the 600VAC meter cc. trol center is supplied from a l
devble-ended bus normally su: plied free a unit lead center l of the 600VAC normal auxiliary poder system and alternately fed from the SSF power system. The motor control center supply bus cannot be connected to either source.unless the cther source breaker is open. Upon L o s 's' o f power to the normal source breaker, an a u t o m a t i c t r a rds f e r to the SSF source will be initiated. Connected to the motor control center are all the remaining 600VAC [oads which require power for hot shutdown of the reactor. Normal power to the pressurizer heaters and necessary primary system iso-Lation valves is supplied from their respective unit normal auxiliary power system. When their normal source arelost, they are transferred to the 600VAC SSF power system.
The 120VAC power cystem consists of a static inverter, a panelboard, a manual transfer switch and interconnecting cables. The system is designed such that normal power is provided from the battery-backed 12SVDC distribution center to the static inverter, through the manual transfer switch and to the panetboard which, in turn, supplies power to the shutdown system circuits. Upon loss of the static inverter or the DC power system the SSF 120VAC power system panelboard is manually transferred to its alternate source, the 600VAC SSF power system.
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s 2.2.3 5.: a r. d b y_ .*.p.s *
- S ur r LY The SSF st andby power su; ply consists of an inde; ndent diesel-ele:tri: ;enerating unit, which is rated at 3000 kW, O.SPF, 4160 VAC The auxiliaries recuired'to assure preper operation of the diesel-generator unit are supplied with power from the appropriate buses (600VAC, 120VAC or 125VDC) of the SSF power system. The diesel-electric
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generating unit is rated for contihucus operation 'at 3000 kW. The design load level for the systet does not exceed the 3000 KW continuous rating of the diesel-electric generating unit. .The unit has.an independent air starting sy st em wi t h st ora ge t o -pr ovide at least tuo slow starts.
conplete with a separate under-An independent fuct system, ground storage tank and a one hour day tank, is supplied The underground for the diesel-electric generating unit.
storage tank is sized to operate.tbe required SSF power The day tank is sized s'ystem for a period of seven days.
based upon the fuel oil storage recuired to successfully start the unit and to allow for orderly shutdown of the diesel unit upon loss of oil kom the main storage tank.
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Following loss of normal (offsite) power, the dies'el-electric generating unit is =anually started and connecte: te the
'"60 VAC
, $$F power system ous. By manuatty closing the 4160VAC generator breaker, the high head auxiliary . service water pump motor breaker, and the 4160/600VAC transformer breaker, the entire SSF power system is provided w.ith its backup scurce of power. .
2.2.4 Q.G._ p. ,e.: _1.ur:tly The DC :ower supply system consists of a 125VDC distribu-tion : enter, one normat and one standby 600VAC/125VDC-battery cha'rger, a 125VDC battery, interconnecting cabtes, and associated instrumentation and control circuits. The
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system is designed to prcvide an adequate and reliable source of continuous DC power for att controls, instrumentation, annunciators, inverters, DC motors, backup lighting, relays ano solenoid vsLves of the SSF pcwer system until the diesel-electric generating unit is available t o s upply cower to the system.
The DC power system is designed to cperate ungrounded and is provided with a ground detection system set to indicate the first ground.
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- Both battery chargerr'areidentical a.c' are rated 600 volts, 3 phase, i: .ert: inpu , ar.d su::ly ;c.er for the 600VAC SSF power system motor control center. N o r m a'l l y ,
one of the chargers is connected to float charge the -
battery while carrying the continuous load,with the other
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charger available to coualize the battery d i s c o n'n e c t e d from the DC bus while the normal charger carries the normal load. It would also be available as a backup for the nor-mal charger. Each charger is designed to prevent the battery from discharging back into any internal c h a r g e.r circuits in the event of an AC power supoly failure or a charger malfunction.
The battery meets the duty cyc!e rec ui renent s wit hout use l
of a charger and without,, decreasing the voltage below an acceptable level in its operating environment. During r.o r m a l operation, the batteries are floated on the buses '
and assume load without interruption upon loss cf a battery charger or AC power source. r i
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s The DC' distribution center receives power 1* e a battery-charger er battery :e:ehcing en AC pewer system sta es, a*.d in turn feecs p we- a DC peser panett are and a' static inverter. The 125VDC distribution centers are metalclad free--standing steel structures of NEMA Type 1A construc-tion with.gasketed doors, cover plates, and contain kolded circuit breakers'and voltage monitoring devices. The
! battery bus voltage is monitored by voltneters located on the 125VDC distribution center.
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3.0 REVIEu maers The design of the S$F was reviewed to the requirements of Appendix R to 10 CFR 50, Sections III.G.3 and III.L, and those requirements applicable for flooding and seismic events.
The Licen ee has stated trat the Standby Shutdown racility (SSF),
associated mechanical and electrical systems and oower su : lies, meets or exceeds tre a: licable cri teria contained 1
in the Oconee FSAR. Additionally, ASME and IEEE codes are utilized as appropriate, in' the oesign of various subsystems and components. The ,SSF and systems / components needed for safe shutdown are designed to withstand the Safe Shutdown ,
Earthquake (SSE).
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The SSF system required for saf t hutdcan are designed with adecuate :apacity to ensure sa'e r. o t shutdown conditions cf a l '. trree 0:: nee units. The SS~ :c.er syster are des'igned with AdeOuate caDacity and capability to supply the necessary loads, and are physically and electrically independent from the main shutdown system power supply (station power). Addi-tionally, the AC and DC power systems and equipment
- required for the SSF essential functions have been designed anc' installed consistent with the Duke QA program of Class 1E ecuipment.
The systems are not designed to meet the single f a i l u r_e criterion, but are designed such that failures in the system do not cause failures or inadvertent operations in existing plant systems. The systems in the SSF are manually initiated; multiple actions must be performed to provide flow to existing safety systems.
4.0 EY:LUATION The SSF is a seiscic Category I structure which houses sy stens and components needed for safe shutdown, and their support systems. The facility is designed to withstand the 1
effects of. flooding and earthquakes. The SSF RC makeup system i has the capability of providing adeouat e makeup to the Reactor Coolant System (RCS), and the capacity of the SSF auxiliary service water system is such that adequate fiow for decay heat removal can be provided to each unit.
- . e o ,
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9 The SSF power supply is physically separated and electrically 4ndepencent thrcugh the transformer and ed :vit b r e a k. e r s free the c a i r. s t a ri'o n power systen. The czpaci ty 21 it.e normal power supply and the diesel generator and the batteries is adequate to supply all the SSF design loads. Adecuate cro-tection is provided to the SSF power supply systems te pro-tect from abnormal e le c t ri ca l condition.
Adecuate instrumentation and controls have been provided in the SSF with the exception of neutron flux and steam generator pressure monitoring capability.
5.0 CONCLUSION
S (with resoect to ftcoc!ine)
Based on our review, we conclude that the design criteria i and the cesign of the power supplies, mechanical syste-s, and instrumentation and controls for the Standby Shutdown Facility with respect to turbine building flooding, are acte: table.
ce"ttusicNS_Ju4+% raenact tn A a m e.ndis R r u u i r - a - =J 6.0 i
6.1 OVERVIEW ,
To preclude a single fire from affecting redundant trains of l.
a system, each safety train must meet minimum separation /
protection requirements. In the licensee's " Fire Hazards 3
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and Response to BTP 9.5-1" dated Decenber 31, 1976, and in subse:uent responses to recuests for additional in :rma-tion, tre ;itensee identifiec vari:;s areas of the plant which cid not meet the recuired separation / protection requirements. To eliminate the deficiencies, a dedicated safe shutdown system independent of the existing plant systems which would be used to achieve safe shutdown was crepcsed. Tc citigate the censecuences of post'ulated fires, the dedicated shutdown system has.been incorporat ed'into the design of the Standby Shutdown Facility (SSF). The use of the SSF as a means of achieving compliance to Sections III.G.3 and III.L of Appendix R to 10 CFR 50 is premis'ed upon the acceptability of the SSF design as a whole. When-
- ever possible, the existing plant systers will be used to achieve hot shutdown. The SSF will be used when the existing plant systems or facilities of one of the three units are unavailable due to a fire, The SSF is not designed to bring the reactor from hot shutdown to cold shutdown. ,
Cold shutdown will be achieved and maintained through the use of existing plant systems and ecuipment as discussed below. No repairs or modifications are required to effect hot shutdown utilizing the dedicated shutdown method. Repairs for cold shutdown may be required dependinc upon the fire area.
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6 .' 2 SYSTEMS USED_FOR POST-FIF.,_; Arc e unt e n oit A. Systems R :uired f o r_,,S a .f.e_ S h u t c an Safe shutdown of the reactor is initially performed by the 1
I insertion of control rods from the control room. Insertion can also be accomplished by removing power to the motor generator sets in the switchgear room. Reactor coolant inventory and reactor shutdown margin are maintai'ned by the SSF makeup pump taking suction from' the spent. fuel pool.
Primary system pressure can be maintained by the pressurizer heaters and pressurizer spray or by use of charging com-bined with Letdown. Should the pressurizer heaters be unavailable (caused by fire inside containment), progres-sion towards cold shutdown will be initiated as soon as hot shutdown is achieved.- Decay heat removal can be accomplished by releasing steam from the steam generators via the atmospheric dump valves. Makeup to the steam generators can be provided by the SSF aunitiary service water system (SSFASWS), which takes suction from the embedded condenser circulating water (CCW) system piping.
Depressurization to cold shutdown can be achieved by bypassing steam to the turbine (since CCW is not lost in the event of a loss of offsite power), use of pressurizer spray, or use e
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i of acxiliary pressurizer spray via the high cressure injec-tion (HPI) pumo. The low p r e s r.u r e injection (LP') oumps wiLL te useo to re.ove decay energy. As a b2ekup to the LPI pumps, the high pressure injection (HPI) pump can be used to maintain flow for decay heat removal. Any damage to either the HPI or LPI power cabling or pump motors can be repaired or replaced within the imposed' time cen-straint to ensure the capability to achieve cold shutdown (pithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the fire.)
Also rtQuired for cold shutdown are the Low pr e s s u r e se r vi c e water (LPSW) pumps. Only one pump per unit is reauired for normal and emergency plant operations. Five LPSW pumps of ecual capacity are provided - three for Units 1 an'd 2, and two for Unit 3. These pumps are separated such that a single fire cannot affect'att pumps. The piping,for these pumps are interconnected so that they may fee'd any of the three units. Any darage to the pump motors or associated power cabling can be repaired, or if necessary, replaced within the imposed time constraint to ensure the capability to achieve cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the fire.
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- 5. Anp Whe f_ Pe.11t J.d _S.b uf.c tia :1_ i. s Ae.;;p L.J 3 The licensee has provided decicated shutdown capab*.'ity in- .
5 ceoen. -t :t tne cabting and e:ui; .ent in t,e c:ntrol room, cable spreading room, and these areas identified in the staff's August 22, 1978 SER on fire protection. The dedi-cated sh'utdown method will be accomplished by'the use of the SSF, and actions performed locally at the equipment.
Electrical isolation will be provided such t'h a t a fire at the SSF will not prevent safe shutdown of the plant'from the control room, and vice versa.
C. Remainino Plant A tp_11 The staff's August 22, 1978 SER identified many areas of the Oconee N u c 'l e a r Station that did not neet various fire protection safe shutdown requirements. Rather than cor-recting the individual deficiencies by modifications to the already installed components, a dedicated shutdown system (the SSF) was prooosed. The i n t e n.t cf the use of the SS? clong with the un amaged systems in the fire 1
affected unit is to neet the requirements d Sections III.G.3 and III.L of Appendix R. l l
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j 6.3 :VALUATI0f1 Perfere:nce Goals .
The. performance 90. L s for pos fire safe s ri u t :: w n can be met using the systems and equipm*nt li st ed in Section A'above.
The control of these functions can be accomplished using the SSF or the centrol room in conj unction with the undamaged systems, in the fire affected unit, depending on the'lpcation of the fire. The transfer of-control :apability between the-control room and the SSF uill be accenplished via a keyed interlock. Annunciation will occur in the control room upon transfer of control.
The process monitoring instruments to be used for a post fire shutdown includes reactor hot leg and cold leg temperatures, reactor coolant pressure and pressurizer level, steam generator level, SSF makeup pump flow and SSF ASWS pump suction and discharge pressure. However, an electrically independent source range flus monitor and steam generator pressure indi-cator hav e no: been p rovi ded. The licensee should be recuested to provide a commitment regarding this instrumentation. The licensee should be advised that the i ns t r um e nt a t i on doe s.*not have to be safety grade, but only meet the requirements of Section III.L.6 of Appendix R.
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- . % g , R The availabLe support systems for post fire safe shutdown are ,
the redundant diesci 'geners ors, vita. buses, and'those support systems associated-with the SSF.
6.4 Esp a is s / 7 2 R o ur. ..R eQ ult Lm.2.n!
Use of the dedicated shutdown method for hot shutdown permits the capability of achieving cold shutd,own within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> af ter a fire event. Repai r's, o r i f ne c es s a ry, replacement of power cabling or pump.. motors associated with LPI, HPI or LPSW may .
be required for cold shutdown. All comp 6nents are stored onsite to ensure the capability to achieve cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the fire. Procedures are available to imple-I ment the requi red repairs / replacements. ,
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6.5 Asso;.iated. Circuits and Isota.tlas AlL circuitry, indicators, instruments and power supplies associated with the SSF are independent from those identified fire zones for which alternative shutdown capability is required. The Licensee has stated that nonsafety related circuits do not run from one redundant train to another and thus negates the possibility of propagating a fire between redundant circuits. The licensee's methods of protecting the safe shutdown capability are consistent with'the guidelines 1 l
r provided by the staff.
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The 5 T instre,entation has a decicate: cower ' source and its cabling is secarated f ror 't hose associated with'the j normal shutdc.n instrurentat'.n. .
All of the normal pows e-d control ri-c.its are :covided i r c '. a t i : n via electri-c' a l l y ec -dinated ci r cuit breakers or fuses. ~
- 2. Cc r.c o n E n c L c s u a The pcwer.rources needed for the SSF ecuipment and instru-entation, and sei t chgea r anc' motor control centers for required components are not located in the postulated fire zones needing alternate shutdown capabilities. Nonsaf,ety related circuits do not run from one redundant t r a i.n t o ,
anotter. Further, all cables of concern are protected by ci r c ui t breakers 4r fuses..._ i
- 3. Sourious Signal operations The devices whose inadvertent operation by spurious signals could adversely affect safe shutdown have been identified.
The cabling required for hot shutdown via the SSF as well as one redundant train of the normal shutdown system are routed to containment through the west penetration rooms for each unit. Cabling for the other independent normal shutdown system is routed to containment through the east penetration rooms. The cast penetration roome are separated at each by n ;hree hour fire barrier. The licen:or.has stated that the cabling routed through the west penetration rooms are separated to the extent practical from existing safety system i
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cabling, and that suitable isstation is previoed between the SSF and existing safety system cabling. Since one indepen-dent shutdown train will be available regardless of a fire in either penetration room, the cable routing design satis-fies the requirements of III.G and is therefore acceptable.
The Licensee has shown that the cable routing of each division (including the SSF cabling) is such that degradation of the redundant shutdown division will not occur, nor will spurious valve actuation occur which might cause an inad-vertent depressurization of the primary in the e v e n't of associated circuit interactions. The cabling for the RHR isolation valves are routed such that the reactor coolant l
pressure boundary integrity will be maintained.
6.6 S.afe Shutdown Procedures and.Manpowe,r The licen,see has committed to develop and implement detailed written procedures for obtaining a safe shutdown condition given a fire event. These procedures will be in place prior to the SSF becoming operational. The manpcwer necessary f'o r accomplishing the operations required f or the alternate shut-down will be available at the plant at all times. Members of the fire brigade will not be included in the shutdown manpower requirements. '
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6.? CONCLUS10::
Based on our teview, we concluded t: the Oconee Nuclear Station
- esi;n .i'. Provide one train o' syste-s necessary to e:hieve and maintain safe shutdown conditions by utilizing either the control room or 't he S$F in conjunction with undamaged systems in the fire,affected unit, and thus will meet the requirements of Appendix R to 10 CFR 50, Sections III.G.3 and III.L with respect to safe shutdown in the event of a fire, with the excep-tions of the availability of a source range flux monitor and steam generator pressure indication at the SSF.
7.0 Overall SSF Conclusion Based on our review, we conclude that the design criteria and the design of the S$F to ensure safe shutdown in t he event of turbine building flooding are acceptable. We also conclude that the Oconee Nuclear Station, in conj unction with the use of the SSF, will meet the requirements of Sections III.G.3 and III.L of Appendix R to 10 CFR 50, with the exception of the availability of a source range flux monitor and steam generator pressure We will require such instrumentation' indication at the SSF.
be provided. This SER also resolves the open items in our SER dated August 22, 1978.
- The ability of the Oconee Nuclear Station in conj unction with l
the SSF to safety shut down following a ><curity incident was t i
l not evaluated. !
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- %, UNITED STATES
[ g NUCLEAR REGULATORY COMMISSION s a wAsumoTom, o. c. zoses FEB 2 71986 h*....
MEMORANDUM FOR: John G. Davis, Director, HMSS Harold R. Denton, Director, NRR ~
i James M. Taylor, Director, IE Thomas E. Murley, Regional Administrator, Region-I -
J. Nelson Grace, Regional Administrator, Region-II James G. Keppler, Regional Administrator, Region-III Robert D. Martin, Regional Administrator, Region-IV John B. Martin, Regional Administrator, Region-V FROM: Victor Stello, Jr.
Acting Executive Director for Operations
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SUBJECT:
MANUAL CHAPTER 0514, MANAGEMENT OF PLANT-SPECIFIC BACKFITTING OF OPERATING POWER REACTORS Enclosed for your implementation is the revised Manual Chapter which has been approved by the Commission. The Manual Chapter incorporates 10 CFR 50.109, 50.54(f), 2.204 and 10 CFR 50, Appendix 0, as they pertain to plant-specific backfitting. The revised Manual Chapter supersedes the draft version which was effective May 1, 1985.
You previously prepared office procedures in accordance with the draf t Manual Chapter. OEDROGR has copies of these procedures and will make the necessary preliminary modifications pursuant to the SRM so they comport to the revised Manual Chapter. They will then be returned to you for final write-up and any other changes you may desire to make. Tom Cox will be interfacing with your staff on this matter over the next few weeks.
The major changes to the Manual Chapter are:
- 1. Addition of procedures for Infor1 nation Requests pursuant to 10 CFR 50.54(f) - MC 1041
- 2. Revised provisions for identification and appeal of staff backfit determinations .MC 1042 and 1044
- 3. Changed provisions of the regulatory ' analysis section to be (cnsistent with provisions in 10 CFR 50.109 - MC 1043
- 4. Defined "Backfitting" consistent with that in 10 CFR 50.109 - MC 1052 The DEOROGR organization will conduct another round of training seminars in the Headquarters and Regional offices to ensure that the staff thoroughly j understands the backfitting process. The schedule for the seminars will be l coordinated with your offices. '
D b b h ? Y -- j f
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lt I If you or your staff have any questions, please contact Jim Sniezek.(492-9704) or Tom Cox (492-4357).
i
[ '
J j
ctor Stell ; Jr.
Acting Executive Di ctor l
for Operations l
Enclosure:
Manual Chapter 0514
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cc: J. Sniezek J. Roe T. Rehm R. Minogue G. Cunningham C. Heltemes
. Norry R. Fraley T. Cox a
. . 1 i
Feb rua ry,1986 l
l MANUAL CHAPTER 1
U.S. NUCLEAR REGULATORY COMMISSION NRC MANUAL Volune: 0000 General Administration 0500 Health and Safety )
Part:
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CHAPTER 0514 NRC PROGRAM FOR MANAGEMENT OF PLANT-SPECIFIC BACKFITTING OF NUCLEAR POWER PLANTS 0514-01 PURPOSE This chapter establishes the requirements and guidance for NRC staff implementation of 10 CFR 50.109 and the provisions of 10 CFR 50 Appendix 0, 10 CFR 50.54(f), and 10 CFR 2.204 as relating to plant-specific backfitting.
Staff requirements and guidance for implementing the provisions of 10 CFR 50.109 pertaining to rules and other generic backfitting are beyond the scope of this Manual Chapter. Pertinent requirements and guidance for generic backfitting are contained in the CRGR Charter. Test and research reactor licensees are not covered by the provisions of this Manual Chapter.
l This chapter defines the, objectives, authorities, and responsibilities and establishes basic requirements for actions to be taken in instances where the NRC staff imposes new plant-specific regulatory staff positions on a nuclear-
-?- February,1986 power plant licensee.I This practice is comonly referred to as "backfitting" and is defined in 10 CFR 50.109 as "the modificatf on of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Comission rules or the imposition of a regulatory staff position interpreting the Consnission rules that is either new or different {
from a previously applicable staff position...." It should be clearly understood j that backfits are expected to occur and are a part of the regulatory process to assure the safety of nuclear power plants. However, it is important for sound and effective regulation that backfitting be conducted in a controlled process.
Plant-specific backfitting is different from generic backfitting in that the fonner involves the imposition on a licensee of positions unique to a particular plant, whereas generic backfitting involves the imposition of the same or similar positions on two or more plants.
The managt. ment of plant-specific backfitting, for which guidance is provided in this document, does not relieve licensees from compliance with the Commission's l
regulations. The management process is intended to provide disciplined NRC review of new or changed positions prior to imposing them at plants without regard to the status of the plant owners efforts in meeting previously applicable requirements or positions which were considered by the staff to provide acceptable levels of safety. The plant-specific backfit management process will enhance l regulatory stability by assuring that changes in regulatory staff positions are in fact required to provide a substantial increase in the overall protection of the public health and safely or common defense and security. Such plant-specific backfitting is entirely proper given the conditions of substantial 1 See Section 05 of this Chapter for a definition of " licensee."
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Februa ry,1986 safety benefit which offset the cost to implement the benefit. This process is intended to provide for effective and efficient use of staff and licensee resources when developing and implementing backfits which enhance nuclear power j l
plant safety.
0514-02 OBJECTIVES It is the overall objective of this program to assure that plant-specific backfitting of nuclear power plants is justified and documented and to specify that the Executive Director for Operations is responsible for the proper implementation of the backfit process. The specific objectives of this program are to allow for substantial improvements in the levels of protection of public health and safety while avoiding any unwarranted burdens on the NRC, public or licensees in implementing backfits. The program should assure to the extent possible that backfits to be issued will in fact contribute effectively and significantly to the health and safety of the public or the common defense and secu rity. This objective is attained by assuring that plant-specific backfits will be communicated to the licensee only after required regulatory analyses are completed and approved as described in Section 0514-042 of this Chapter and that the backfit and supporting regulatory analyses are approved by the appropriate Office Director or Deputy Director, or Regional Administrator or Deputy Regional Administrator and forwarded to the Executive Director for Operations before the backfit and supporting analysis are communicated to the licensee.
This Manual Chapter governs those plant-specific backfits communicated to the licensees or identified by licensees after May 1, 1985.
0514-03 RESPONSIBILITIES AND AUTHORITIES 031 The EDO is responsible to the Commission for plant-specific backfit actions. Office Directors and Regional Administrators have the authority I
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- .4. February,1986 to review cases and make decisions fo:- the agency on individual actions as described in other sections of this Manual Chapter. The E00 may review and modify any plant specific backfit decision at his or her initiative or at the request of a licensee in accordance with Section 044 The EDO may authorize deviations from this Chapter when the E00 finds that such action is in the public interest and the deviation otherwise complies with the applicable regulations.
032 The Director, Regional Operations and Generic Requirements (ROGR) Staff, shall assure that process controls for overall agency management of the plant-specific backfit process are developed and maintained. These process controls shall include specific procedures, training, progress (
monitoring systems, and provisions for obtaining and evaluating both staff and industry views on the conduct of the backfit process. The Director, ROGR Staff, is also responsible for assuring that each licensee is informed of the existence and structure of the NRC program described in this Manual Chapter. The Director, ROGR Staff, shall assure that i
substantive changes in the Manual Chapter and related precedures are communicated to the licensees. J 033 The Director, Office of Nuclear Reactor Regulation (NRR) shall assure that an overall procedure for managing plant-specific backfitting that involves positions taken by NRR is developed, implemented, and maintained, in accordance with this Manual Chapter. The overall procedure will be approved by the E00. The Director, NRR, shall consult and coordinate with Regional Administrators and the Directors of the Office of Inspection and Enforcement and the Office of Nuclear Materials Safety and Safeguards, as '
appropriate, to develop resolutions of proposed plant-specific backfits in program areas for which NRR has responsibility.
For backfits within NRR's program area of responsibility which are l proposed by HRR staff, the Director or Deputy Director, NRR, without further delegation, shall approve the regulatory analysis prior to l ,
)
February,1986 connunicating the backfit and analysis to the licensee. For backfits within the NRR program area of responsibility, but which are proposed by ,
other NRC staff who have been delegated NRR program implementation and decision authority, regulatory analysis shall be approved by the staff )
Office Director / Administrator or Deputy Director / Administrator of the NRC staff person proposing the backfit, prior to connunicating the backfit and analysis tc the licensee. For all backfits within the NRR program area of responsi'.Mity which are appealed by a licensee, the Director, NRR, shall make the decision on imposition of the backfit. The decision is subject to EDO review under Section 031. The Director, NRR, shall assure NRR staff performance in accordance with this Manual Chapter. I f
934 The Director, Office of Inspection and Enforcement (IE) shall assure that an overall procedure for managing punt-cpecific backfitting that involves positions taken by IE is developed, implemented, and maintained, in accordance with this Manual Chapter. The overall procedure shall be approved by the EDO. The Director, IE, shall consult and coordinate with Regional Administrators and the Directors of Nuclear Reactor Regulation and Nuclear Material Safety and Safeguards as appropriate, to develop resolutions of proposed plant-specific backfits in program areas for which j
!! has responsibility.
I for backfits within the IE program area of responsibility which are proposed by IE staff, the Director or Deputy Director, IE, without further ,
delegation, shall approve the regulatory analysis prior to connunicating the backfit and analysis to the licensee. For backfits within the IE l program area of responsibility, but which are proposed by other NRC staff ;
who have been delegated IE program implementation and decf sior. authority, the regulatory analysis shall be approved by the staff Office Director /
Administrator or Deputy Director / Administrator of the NRC staff person proposing the backfit, prior to connunicating the backfit and analysis to the licensee. For all beckfits within the IE program area of responsibility I
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i February,1986 ;
which are appealed by the licensee, the Director, IE, shall make the decisicn on imposition of the backfit. The decision is subject to EDO review under Section 031. The Director, IE, shall assure IE staff l performance in accordance with this Manual Chapter.
035 The Director, Office of Nucle:r Material Safety and Safeguards (NMSS),
shall assure that an overall procedure for managing plant-specific backfitting that involves positions taken by NMSS is developed, imple-mented, and maintained, in accordance with this Manual Chapter. The overall procedure shall be approved by the EDO. The Director, NMSS, shall consult and coordinate with Regional Administrators and the Directors of j the Offices of L clear Reactor Regulation and Inspection and Enforcement, l as appropriate, to develop resolutions of proposed plant-specific backfits in program areas for which NMSS has responsibility.
1 i
For backfits within the NMSS program area of responsibility which are proposed by NMSS staff, the Director or Deputy Director NMSS, without further delegation, shall approve the regulatory analysis prior to communicating the backfit and analysis to the licensee. For backfits within the NMSS program area of responsibility, but which are proposed by other NRC staff who have been delegated NMSS program implementation end decision authority, the regulatory analysis shall be approved by the staff Offhe Director / Administrator or Deputy Director / Administrator of the NRC staff person proposing the backfit, prior to consnunicating the backfit and analysis te Se licensee. For all backfits within the HMSS program area of responsibility which are appealed by a licensee, the Director, HMSS, shall make the decision on imposition of the backfit. The decision is subject to EDO review under Section 031. The Director, NMSS, shall assure HMSS staff perfonnance in accordance with this Manual Chapter. i 036 The Regional Administrator of each region shall assure that an overall procedure for managing plant-specific backfitting that involves positions taken by the region in any program area for which the region nas been delegated authority, is developed, implemented, and maintained, in
e f
February,1986 accordance with this Manual Chapter. The overall procedure shall be approved by the EDO.
The Regional Administrator of each Region shall consult and coordinate with the Directors of the Offices of Nuclear Reactor Regulation, Inspection and Enforcement, and Nuclear Material Safety and Safeguards as appropriate, to identify issues and develop resolutions of proposed plant-specific backfits where such backfitting would result from positions taken by the Region.
For backfits proposed by the Region, the Regional Administrator or Deputy pegional Administrator, without further delegation, shall approve the regulatory analysis prior to communicating the backfit and analysis to the licensee. For backfits proposed by the Region and appealed by the licensee, the Administrator is responsible for the conduct of the appeal process within the region; however, if agreement cannot be reached at the regional level, the decision on imposition of the backfit shall be made by the Director of the program office having responsibility for the program area relevant to the backfit. The decision is subject to EDO review under Section 031. Each Regional Administrator shall assure Regional staff performance in accordance with this Manual Chapter.
i 037 The Director, Office of Resource Management shall, in coordination with the Director, Regional Operations and Generic Requirements Staff, the Office Directors, and Regional Administrators, develop and maintain the overall NRC data base management system identified and described in Section 046 of this Chaoter.
038 NRC staff positions may be identified as potential backfits either by HRC staff or by persons who are not members of the NRC staff. Such identifications will be considered by the Office Director / Administrator having responsibility to develop staff positions on the matter at issue.
This Office Director / Administrator will be responsible to make the
February,1986 determination as to whether the staff position is a backfit. If the staff position is determined to be a backfit, a regulatory analysis shall be completed prior to connunicating the backfit and analysis to the licensee and forwarding to the EDO. If the staff position is determined not to be a backfit, a documented evaluation will be completed and the Office Director / Administrator having the responsibility to develop a staff position on the issue will deal with the matter in accordance with appropriate office procedures, i
0514-04 BASIC REQUIREMENTS 041 Information Requests Pursuant to 10 CFR 50.54(f)
A revision to 10 CFR 50.54(f) was issued with the September 20, 1085 revision to 10 CFR 50.109 in the Federal Register (50 8 38097). The revision ger.erally requires that the NRC prepare reasons for issuing infonnation requests prior to issuance. Concerning the review of appli-cations for licenses or amendments, or the conduct of inspection activi-ties, for plants under construction, no analysis will be necessary if
- t. staff seeks information of a type routinely sought as a part of the standard procedures applicable to the review of applications. If the request is not part of routine licensing review, for example, it seeks to gather infonnation pursuant to development of a new staff position, then a Staff analysis of the reasons for the request must be prepared and approved prior to issuance.
Concerning licensing review or inspection activities for operaf ing plants, only information requests seeking to verify licensee compliance with the current licensing basis for the facility are exempt from the necessity to prepare the reason or reasons for the request. Requests for infonnation to detennine compliance with existing facility require- j ments including fact-finding reviews, inspections and investigations of accidents or incidents, usually are not made pursuant to Section 50.54(f), l l
l
(
February,1986 j
nor are- such requests normally considered within' the scope of the backfit .
rule or this Manual Chapter.
The Directors of NRR, IE and NMSS and the Regional Administrators shall develop internal office procedures to ensure that there is a rational basis for all infonnation requests not clearly excepted from evaluation, whether or not it is clear that backfit action would result from staff evaluation of the information supplied by the licensee. The request must be evaluated to determine whether the burden imposed by the.inforntion request is justified in view of the potential safety significance of the issue to be addressed. The information request
! and the staff evaluation must be approved by the cognirant Office Director or Regional Administrator prior to transmittal of the request for information to a licensee.
NRC staff evaluations of the necessity for an information request shall include at least the following elements:
- 1. A problem statement that describes the need for the information in terms of potential safety benefit.
P. The licensee actions required and the cost to develop a response to the information reauest.
- 3. An anticipated schedule for NRC use of the information.
042 Identifying Plant-Specific Backfits The NRC staff shall be responsible for identifying proposed plant-specific backfits as defined by Section 05 of this Manual Chapter. It is expected that staff at all levels will evaluate any proposed plant-specific posi-tion with respect to whether or not the position qualifies as a proposed
( backfit pursuant to Section 05 of this Manual Chapter. No staff position
__m___________.___._. . . _
=
Februa ry,1986 should be conynunicated to a licensee unless the NRC official communicating that position has ascertained whether or not the position is to be identified as a backfit. Appendix A to this Manual Chapter provides additional information to help in identifying backfits arising from selected staff activities. When a staff proposed position is identified as a backfit and imposition of the backfit is not necessary to ensure that the facility poses no undue risk to public health and safety, the appropriate staff office should proceed promptly with the preparation of a regulatory analysis (Section 043) for approval in accordance with this Chapter. The staff may, at any point in the development of the regulatory analysis, decide that further analysis is likely to show either that the proposed safety benefit f s not likely to be substantial additional overall protec-tion, or that the direct and indirect costs of implementation are not likely to be justified. In this case, the issue may be closed, with appropriate notice sent to all parties and recorded in the recordkeeping system described in Section 046.
When (1) a staff proposed position is determined not to be a backfit because the proposed modification is necessary to bring a facility into compliance with a license or the rules or orders of the Comission, (Sections 052-1,053-1) or into confonnance with written commitments by thelicensee(Sections 052-1, 053-2), or (2) the Director of NRR, HMSS or IE detennines that imposition of a backfit is necessary to ensure that the facility poses no undue risk to public health and safety, no regulatory analysis is required. Instead, the appropriate Director is to provide a documented evaluation to support the action taken. The evaluation shall include a statement of the objectives of and reasons for the modification and the basis for invoking the exception. In the case of a backfit needed to assure that the facility poses no undue risk to public health and safety, the documented evaluation shall also include an analysis to document the safety significance and appropriateness of the action taken and consideration of how costs contribute to selecting the solution among various acceptable alternatives. Such an evaluation
s February,1986 is to be issued with the backfit except that, when an imediately effec-tive regulatory action is necessary, and the safety need is so urgent that full documentation cannot be completed, the documentation may follow the backfit.
A proposed staff position which is not identified by the NRC staff as a backfit position may be claimed to be a backfit position by a licensee.
The staff will promptly consider a licensee claim of backfit to determine if the claimed backfit qualifies as such in accordance with Section 05 of this Chapter. Licensees identifying such items should send a written claim of backfit (with appropriate supporting rationale) to the Office Director or Regional Administrator of the NRC staff person who issued the position with a copy to the EDO. If the NRC staff detennination is that the issue is a backfit, the appropriate staff office should proceed immediately with the preparation of the regulatory analysis for approval in accordance with this Chapter.
If the determination is that the proposed staff position is not a backfit, the appropriate staff office shall document the basis for the decision and transmit it together with any documented evaluation required by this section to the licensee. In any case, the appropriate Office Director /
Regional Administrator shall report to the ED0 and inform the licensee, within 3 weeks after receipt of the written backfit claim, of the results of the determination and the plan for resolving the issue. ,
l When a licensee is informed that a claimed backfit is, in the iudgment of the NRC, not a backfit, the licensee may appeal this determination as described in Section 044-of this Chapter.
043 Regulatory Analysis Positions identified as plant-specific backfits requiring the analysis in this section shall be transmitted to licensees only after a !
1
February,1986 i
detennination, that there is a substantial increase in the overall protec-tion of'the public health and safety or the comon defense and security to be derived from the backfit, and that the direct and indirect costs of implementation for that facility are justified in view of the increased protection. The proposed backfit and supporting regulatory analysis must be approved by the appropriate Program Office Director or Deputy Director, or Regional Administrator or Deputy Regional Administrator and fondarded' to the EDO before the backfit and its supporting regulatory analysis are transmitted to the licensee.
The regulatory analysis shall generally conform to the directives and L guidance of NUREG/BR-0058 and NUREG/CR-3568, which are the NRC's governing documents concerning the need for and preparation of regulatory analyses, in preparing regulatory analyses under this section..the staff should note that the complexity and comprehensiveness of an analysis should be limited to that necessary to provide.sn adequate basis for decisionmaking among the alternatives available. The emphasis should be on simplicity, flexi-bility, and comon sense, both in terms of the type of information supplied and in the level of detail provided. The following infonnation and any other infomation relevant and material to the backfit shall be included in the regulatory analysis, as available and appropriate to the analysis:
- 1. A statement of the specific objective that the proposed backfit is designed to achieve. This should also include a succinct description of the backfit proposed, and how it provides a substantial increase in overall protection.
- 2. A general description of the activity that would be required by the licensee in order to complete the backfit.
- 3. The potential safety impact of changes in plant or operational comp.lexity, including the relationship to proposed and existing regulatory requirements.
Februa ry,1986 4 Whether the proposed backfit is interim or final and, if interim, the justification for imposing the proposed backfit on an interim basis.
- 5. A statement that describes the benefits to be achieved and the cost to be incurred. This statement should include consideration of at least the following listed factors. Information should be used to the extent that it is reasonably available, and a qualitative assessment of benefits may be made in lieu of the quantitative analysis where it would provide more meaningful insights, or is the only analysis practicable,
- a. The potential change in risk to the public from the accidental offsite release of radioactive material.
- b. The potential impact on radiological exposure of facility employees.
Also consider the effects on other onsite workers, due both to installation of procedural or hardware changes and to the effects of the changes, for the remaining lifetime of the plant.
- c. The installation and continuing costs associated with the backfit, including the cost of facility downtime or the cost of construction delay.
- d. The estimated resource burden on the NRC associated with the proposed backfit and the availability of such resources.
- 6. A consideration of important qualitative factors bearing on the need for the backfit at the particular facility, such as, but not limited to, operational trends, significant plant events, management effectiveness, or results of performance reports such as the Systematic Assessment of Licensee Performance.
- 7. A statement affirming appropriate interoffice coordination related to the proposed backfit and the plan for implementation, l
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Februa ry,1986 j
- 8. The basis for requiring or permitting implementation on a particular schedule, including sufficient information to demonstrate that the i schedules are realistic and provide adequate time for in-depth engineering, evaluation, design, procurement, installation, testing, development of operating procedures, and training of operators and other plant personnel, as appropriate. For those plants with approved integrated schedules, the integrated scheduling process can be used for implementing this step and the following two procedural steps. l
- 9. A schedule for staff actions involved in implementation and verification of implementation of the backfit, as appropriate.
- 10. Importance of the proposed backfit considered in light of other safety-related activities unde may at the affected facility.
044 Appeal Process The appeal processes described in this section are of two types, applied to two distinctly different situations:
- 1. Appeal to an Office / Region to modify or withdraw a proposed backfit which has been identified, and for which a regulatory analysis has been prepared and transmitted to the licensee; or
- 2. Appeal to an Office / Region to reverse a denial of a prior licensee claim that a staff position, not identified by the NRC as a backfit, requiring a regulatory analysis as described in Section 043 is such a proposed backfit.
After a backfit has been identified, or claimed and then verified to be a backfit, in accordance with this Chapter (see Section 042), and a licensee has been notified of staff intent to impose a backfit, a licensee may choose to appeal the proposed plant-specific backfit to Npt staff i - - - - - _ _ _ _ _ _ _
i
- Februa ry,1986 management to request that the proposed backfit be withdrawn or modified.
Licensees will address an appeal of a proposed backfit to the Office Director or Regional Administrator whose staff proposed the backfit with a copy to the EDO. The appeal should provide arguments against the rationale for imposing a backfit as presented in the staff's regulatory analysis.
The Office Director or Regional Administrator shall report to the EDO within 3 weeks after receipt of the appeal concerning the plan.for resolv-ing the issue. The licensee should also be promptly and periodically informed in writing regarding the staff plans.
The decision of the Office Director on an appeal of plant-specific backfit may be appealed to the EDO unless resolution is achieved at a lower management level. The EDO shall promptly resolve the appeal and shall state his reasons therefor. Summaries of all appeal meetings shall be prepared promptly, provided to the licensee, and placed in appropriate Public Document Rooms. During the appeal process, primary consideration shall be given to how and why the proposed backfit provides a substantial increase in overall protection and whether the associated costs of implementation are justified in view of the increased protection. This consideration should be made in the context of the regulatory analysis as well as any other information that is relevant and material to the proposed backfit.
l After a proposed staff position has been claimed by a licensee to be a backfit position requiring a regulatory analysis and then determined by the NRC not to be such a backfit, the licensee may appeal the NRC decisior. l regarding the backfit determination. The appeal should be addressed to, and will be decided by, the Director of the program office having responsi bility for the program area relevant to the staff position, unless resolu-tion is achieved at a lowe- management level. A copy of the appeal should also be sent to the Executive Director for Operations. The appeal should take into account the staff's documented evaluation, the licensee's response, and any other information that is relevant and material to the backfit
February,1986 I I
determination. The EDO may review and may modify a decision either at j his or her own initiative or at the request of the licensee, if the licensee appeals to the EDO, the EDO shall promptly resolve the appeal and shall state the reasons therefor. Backfit claims and resultant staff q determinations that are reevaluated in response to an appeal, and that are again determined by the NRC not to be backfits, are not to be treated '
further in the context of this Manual Chapter. Such matters are to be dealt with within the normal licensing or inspection appeal process and are not subject to the requirements of this Manual Chapter.
^
045. Implementation of Backfits Following approval of the regulatory analysis by the appropriate Office Director or Regional Administrator, review if any by the EDO, and issuance of the backfit to the licensee, the licensee will either implement the backfit or appeal it. Af ter an appeal and subsequent final decision by the appropriate Office Director or FDO, the licensee may elect to imple-ment a backfit resulting from the decision. if the licensee does not elect to implement the backfit, it may be imposed by Order of the appropriate Office Director.2 Implementation of plant-specific backfits will normally be accomplished on a schedule negotiated between the licensee and the NRC. Scheduling criteria should include the importance of the backfit relative to other safety related activities underway, or the plant construction or mainten-ance planned for the facility, in order to maintain high quality construc-tion and operations. For plants that have integrated schedules, the integrated scheduling process can be used for this purpose.
I 2 Once an Order is issued, whether or not it is immediately effective, this Chapter no longer applies and appeals are governed by the procedures in 10 CFR Part 2, Subpart B.
4 _ . . . . . .
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Februa ry,1986 )
3 A staff-proposed backfit may be imposed by Order prior to completing any of. the procedures set forth in this Manual Chapter provided the NRC official authorizing the Order determines that imediate imposition is necessary to insure that the facility poses no undue risk to the public health and safety or the consnon defense and security. In such cases, the E00 shall be notified promptly of the action and a documented evaluation as described in Section 042 perforned, if possible, in time to be 13 sued with the order. l If "insnediate imposition" is not necessary, staff proposed backfits shall not be imposed, and plant construction, licensing action, or operation ]
shall not be interrupted or delayed by NRC actions, during the staff's evaluation and backfit transmittal process, or a subsequent appeal )
, process, until final action is completed under this Chapter.
046 Recordkeeping and Reporting The proposing Headquarters Office or Regional Office shall adminis-tratively manage each proposed plant-specific backfit using a record-keeping system that provides for prompt retri'e val of current status, planned and accomplished schedules, and ultimate disposition. Office systems shall be compatible with an overall NRC data base accessible to appropriate NRC managers. The systems shall provide reference to all documents issued or received by NRC staff relative to a plant-specific )
backfit, including requests, positions, statements, and sumary reports.
Specific data required will include, but are not limited to:
I 3 Refer to Footnote 2 i
h l
1
i February, 1986 r
- 1. Licensee and facility affected.
- 2. Whether a backfit is identified by staff or by a licensee, t
- 3. Identification and description of the document that either transmits a staff-identified backfit or a licensee request for consideration of a licensee-identified backfit.
4 Substance of the backfit issue.
- 5. In the case of a licensee-identified backfit, the dates (predicted and completed) that determinations are made as to whether or not a staff position qualifies as a backfit, the substance of the deter-mination, and the organization and official responsible to make the determination.
- 6. A brief description of what action is pending, and the officials responsible to complete the action.
- 7. Action closing data, to include a description of licensee or staff action and date of agreement or order to implement; responsible officials and organization for each action.
047 Exceptions Nothing in this Manual Chapter shall be interpreted as authorizing or requiring the staff to make plant-specific backfits or assessments for generic backfits that are, or have been, subject to review by the CRGR cad approval by the EDO, or for generic backfits approved prior to November 1981, unless the EDO determines that significant plant-specific issues I
i I
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' Februa ry,1986 i
were not considered during the prior reviews or the EDO authorizes a deviation under Section 031.
048 References
- 1. NUREG/BP-0058, Rev. 1, May 1984, " Regulatory Analysis Guidelines of the U. S. Nuclear Regulatory Comission"
- 2. NUREG/CR-3568, December 1983, "A Handbook for Value-Impact Assessment" j
- 3. NUREG/CR-3971, October 1984, "A Handbook for Cost Estimating" i
- 4. Revision of Backfit Rule, Code of Federal Regulations, 50 FR 38097 (Sept. 20, 1985) 0514 05 DEFINITIONS 051 Licensee Except where defined otherwise, the word licensee as used in this Manual Chapter shall mean that person that holds a license to operate a nuclear power plant, or a construction permit to build a nuclear power plant, or a j
Preliminary Design Approval or Final Design Approval for a Standardized Plant Design.
052 plant-Specific Backfit Backfitting is defined as the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization i
required to design, construct or operate a facility; any of which may result from a new or amended provision in the Comission rules or the i
j
)
I
Februa ry,19P6 imposition of a regulatory staff position interpreting the Curission rules that is either new or different from a previously applic.?ble staff position after certain specified dates. Backfitting is " plant-specific" when it involves the imposition of a position that is unique to a particular plant.
It should be noted that to be a plant-specific backfit a staff position must meet conditions involving both (1) the substance of the elements of a proposed staff position and (2) the time of the identification of the staff position:
- 1. A staff position may be a proposed backfit if it would cause a licensee to change the design, construction or operation of a facility from that consistent with already applicable regulatory staff positions. Applicable regulatory staff positions are described in Section 053.
- 2. A staff position as described in (1) above is a proposed backfit if it is first identified to the licensee after certain important design, construction or operation milestones, involving NRC approYals of varying kind, has been achieved. Those times after which a j new or revised staff position will be considered a backfit are as follows:
- a. after the date of issuance of the construction permit for the facility (for facilities having construction permits issued afterMay1,1985);or
- b. after 6 months before the date of docketing of the OL application for the facility (for facilities having construction permits issued before May 1, 1985); or
Feb ruary,1986 l
- c. after the date of issuance of the operating license for the facility (for facilities having an operating license on May 1, 1985).
- d. After the date of issuance of the design approval under 10 CFR 50, Appendix M, N or 0.
NOTE: 10 CFR 50.109 was revised and issued on September 20, 1985 (50 FR 38097). MC-0514 was implemented on May 1,1985. The current revision of the Backfit Rule 50.109 states that paragraph 50.109(a) of the rule does not apply to backfits imposed prior to the effective date of the rule. However, the EDO directives embodied in this Manual Chapter-0514 have been effective and remain effective as of May 1, 1985.
053 Applicable R,egulatory Staff positions Applicable regulatory staff positions are those already specifically imposed upon or committed to by a licensee at the time of the identifica-tion of a plant-specific backfit, and are of several different types and sources:
- 1. Legal requirements such as in explicit regulations, orders, plant licenses (amendments, conditions, technical specifications). Note that some regulations have update features built in, as for example, 10 CFR 50.55a, Codes and Standards. Such update requirements are applicable as described in the regulation.
- 2. Written commitments such as contained in the FSAR, LERs, and docketed correspondence, including responses to IE Bulletins, responses to Generic Letters, Confirmatory Action Letters, responses to inspection Reports, or responses to Notices of Violation.
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- 3. NRC staff positions that are documented, approved, explicit interpre-tations of the more general regulations, and are contained in documents such as the SRP, Branch Technical Positions, Regulatory Guides, Generic Letters and IE Bulletins and to which a licensee or an applicant has previously comitted to or relied upon. Positions contained in these documents are not considered applicable staff positions to the extent that staff has, in a previous licensing or inspection action, tacitly or explicitly excepted the licensee from part or all of the position.5 0514-06 APPENDIX A - Guidance tor Making Backfit Determinations A. General In this section selected regulatory activities and documents are discussed in order to enable members of the NRC staff and the regulated industry to )
better understand the conditions under which a staff position may be viewed as a plant-specific backfit. It is important to understand that the necessity for making backfit determinations should not inhibit the nor:r.al informal dialogue between the technical reviewer or inspector and the licensee. The intent of this process is to manage backfit imposition, not to quell it. The discussion in this Appendix is intended to aid in identifying backfits in accordance with the principles and the practices that should be implemented by all staff members. This Appandix is not intended to be an exhaustive, comprehensive workbook in which can be found a parallel example for each situation that may arise. As is evident from the definitions in Section 05 of this Manual Chapter, a plant-specific )
l l
4 Requirements may be imposed by rule or order. Staff interpretations I such as examples of acceptable ways to meet requirements are not requirements in and of themselves, l
5 Imposition of a staff position from which a licensee has previously been i excepted is a backfit. !
4
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- ?3 - Feb rua ry,1986 backfit has the elements of a change from an already applicable staff position where an applicable staff position is defined as that established before certain defined milestones in the affected plant's licensing history. There will be some judgment necessary to determine whether a staff position would cause a licensee to change the design, construction or operation of a facility. In making this determination, the fundamental cuestion is whether the staff's action is directing, telling, or coercing, or is merely suggesting or asking the licensee to consider a staff proposed action.
Actions proposed by the licensee are not backfits under this chapter even though such actions may result from normal discussions between staff and licensee concerning an issue, and even though the change or additions may meet the definitions of Section 052 and 053.
B. Licensing
- 1. Standard Review Plan (SRP) - The SRP delineates the scope and depth of staff review of licensee submittals associated with various licensing activities. It is a definitive NRC staff interpretation of measures which, if taken, will satisfy the requirements of the more generally stated, legally binding body of regulations, primarily found in title 10 CFR. Since October 1981, changes to the SRP are to j have been reviewed and approved through a generic review process j involving the Committee to Review Generic Requirements (CRGR), and the extent to which the changes apply to classes of plants is defined. Consequently, application of a current SRP in a specific operating license (OL) review generally is not a plant-specific backfit, provided the SRP was effective 6 months prior to the start of the OL review. Asking an applicant for an operating license questions to clarify staff understanding of proposed actions, in order to determine whether the actions will meet the intent of the i SRP, is not considered a backfit.
l
Februa ry,1986 On the other hand, using acceptance criteria more strinoent than those contained in the SRP or taking positions more stringent than or in addition to those specified in the SRP, whether in writing or orally, is a plant-specific backfit. During meetings with the licensee, staff discussion or coments regarding issues and licensee actions volunteered which are in excess of the criteria in the SRP generally do not constitute p16nt-specific backfits; however, if the staff implies or suggests that a specific action in excess of already applicable staff positions is the only way for the staff to be satisfied, the action is considered a plant-specific backfit whether or not the licensee agrees to take such action. However, the staff
- hould recognize that a verbally implied or suggested action should not be accepted by a licensee as an NRC position of any kind, backfit or not; only written and authoritatively approved position statements should be taken as NRC positions.
l l Application of an SRP to an operating plant after the license is granted generally is considered a backfit unless the SRP was approved specifically for operating plant implementation and is applicable to l
such operating plant. It is important to note, however, that in order to issue an amendment to a license, there must be a current finding of compliance with regulations applicable to the amendment.
As a specific example, review of a plant owner's application for a license amendment to authorize installation and operation of a new reactor core, comonly called a " reload application," may necessitate review of new fuel designs or new thennal-hydraulic correlations and associated operating limits. Such changes that are clearly advances in design or operation may involve new or unreviewed safety issues, and may warrant review to SRP criteria which were approved subsequent to initial license issuance to the licensee. This is not considered a backfit. However, such review to newer SRP revisions is not neces-sarily reouired to determine current compliance with regulations.
Licensee proposed revisions in design or operation that raise staff
{
February, 1986 questions only about potential reduced margins of safety as defined in the basis for any technical specification should be reviewed by reanalysis of the same accident sequences and associated assumptions as analyzed in the FSAR for the initial license issuance.
During relcad reviews, staff proposed positions with regard to technical matters not related to the changes proposed by a licensee shall be considered backfits.
?. Regulatory Guides - As part of the generic review process pursuant to the CRGR Charter, it is decided which plants or groups of plants should be affected by new or modified Regulatory Guide provisions.
Such implementation is therefore not governed by the plant-specific backfit procedures. However, any staff proposed plant-specific implementation of a Regulatory Guide provision, whether orally or in writing, for a plant not encompassed by the generic implementation determination is considered a plant-specific backfit. A staff action with respect to a specific licensee that expands on, adds to, or modifies a generically approved regulatory guide, such that the position taken is more demanding than intended in the generic positions, is a plant-specific backfit.
- 3. Plant-Specific Orders - An order issued to cause a licensee to take actions which are not otherwise applicable regulatory staff positions is a plant-specific backfit. As described in Section 044 of this Manual Chapter, an order effecting prompt imposition of a backfit may be issued prior to completing any of the procedures set forth in this Manual Chapter provided that the appropriate Office Director l
determines that prompt imposition is necessary, I'
An order issued to confirm a licensee comitment to take specific l action even if that action is in excess of previously applicable l
L-_-__--___ _
4 February, 1986 staff positions, is not a plant-specific backfit provided the connit-ment was not obtained by the staff with the expressed or implied direction that such a connitment was necessary to gain acceptance in the staff review process. Discussion or connents by the NRC staff identifying deficiencies observed, whether in meetings or written reports, do not constitute backfits. Definitive statements to the licensee directing a specific action to satisfy staff positions are backfits unless the action is an explicit already applicable regulatory i staff position.
C. Inspection and Enforcement 4
- 1. Inspections - NRC inspection procedures are to govern the scope and depth of staff inspections associated with licensee activities such as design, construction and operation. As such, they define those items the staff is to consider in its determination of whether the licensee is conducting its activities in a safe manner. The conduct )
of inspection establishes no new staff positions for the licensee and l
1s not a plant-specific backfit.
]
J l
Staff statements to the licensee that the contents of an NRC inspec-tion procedure are positions that must be met by the licensee constitute .
a plant-specific backfit unless the item is an applicable regulatory '
staff position. Discussion or connent by the NRC staff regarding deficiencies observed in the licensee conduct of activities, whether in meetings or in written inspection reports, do not constitute back-fits, unless the staff suggests that specific corrective actions different from previous applicable regulatory staff positions are the only way to satisfy the staff. In the nomal course of inspecting to l
detemine whether the licensee's activities are being conducted safely, inspectors may examine and make findings in specific technical areas wherein prior NRC positions and licensee commitments do not i exist. Examination of such areas and making findings is not considered I __ _ _ _ _--_ _ .-.
. February,1986 a backfit. Likewise, discussion of findings with the licensee is nnt considered a backfit. If during such discussions, the licensee agrees that it is appropriate to take action in response to the inspector's findings, such action is not a backfit provided the inspector does not indicate that the. specific actions are the only way to satisfy the staff. On the other hand, if the inspector indicates that a specific action must be taken, such action is a backfit unless it constitutes an applicable regulatory staff position. Further, if the licensee decides to claim that the inspector's findings are a backfit, then the staff must decide whether they are a backfit under this Chapter.
For example, if the licensee commits to ANSI-N18.7 in the SAR and the inspector finds the licensee's implementing procedures do not contain all the elements required by ANSI-N18.7 telling the licensee he must take action to include all the elements in the implementing procedures is not a backfit. If the inspector finds the licensee has included all the required elements of ANSI-N18.7, but has not included certain '
of the optional elements in th'e implementing procedures, inspector discussion with the licensee regarding the merits of including the optional elements is not a backfit. On the other hand, if the inspector tells the licensee that the implementing procedures must include any or all of the optional elements in order to satisfy tha staff, inclusion of such elements is a backfit, whether or not agreed to by the licensee.
P. Notice of Violations (NOV) - a NOV requesting description of a licensee's proposed corrective action is not a backfit. The licensee's comitments in the description of corrective action are not backfits. A request by the staff for the licensee to consider some specific action in response to an NOV is not a backfit. However, I if the staff is not satisfied with the licensee's proposed corrective actions and requests that the licensee take additional actions, those
. Feb rua ry,1986 additional actions (whether requested orally or in writing) are a backfit unless they are an applicable regulatory staff position.
Discussions during enforcement conferences and responses to the licensees requests for advice regarding corrective actions are not backfits; however, definitive statements to the licensee directing a specific action to satisfy staff positions are backfits, unless the action is an explicit applicable regulatory staff position.
- 3. Billetins - IE Bulletins and resultant actions requested of licensees undergo the generic review process pursuant to the CRGR Charter.
Therefore, in general, it is not necessary to apply the plant-specific backfit process to the actions requested in a Bulletin.
However, if the staff expands the action requested by a Bulletin during its application to a specific licensee, such expansion is considered a plant-specific backfit.
4 Reanalysis of Issues - Throughout plant lifetime, many individuals on the NRC staff have an opportunity to review the requirements and commitments incumbent upon a licensee, Undoubtedly, there will be occasions when a reviewer concludes the licensee's program in a specific area does not satisfy a regulation, license condition or connitment. In the case where the staff previously accepted the licensee's program as adequate, any staff specified change in the program would be classified as a backfit.
For example, in the case of an NTOL, once the SER is issued signifying staff acceptance of the programs described in the SAR, the licensee should be able to conclude that his commitments in the SAR satisfy the flRC requirements for a particular area. If the staff was to subsequently require that the licensee commit to additional action other than that specified in the SAR for the particular area, such action would constitute a backfit.
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4 A somewhat different situation exists when the licensee has made a submittal committing to a specific course of action to meet an applicable position, and the staff has not yet responded, and there-fore has not indicated that the comitment is or is not sufficient to meet the applicable position. Subsequent staff action, which must be taken within a reasonable time not delaying the applicants' implemen-tation plans, to cause the licensee to meet the applicable regulatory staff position is not a backfit. If the licensee has moved aread in the intervening time to implement that which the licensee propnsed to do in its submittal and the staff has failed to provide a timely response, then the staff position may be considered a backfit. Thus, if a licensee has implemented a technical resolution intended to meet an applicable regulatory staff position, and staff for an extended period simply allows the licensee resolution to stand with tacit acceptance indicated by nonaction on the part of NRC, then a subsequent action to change the licensee's design, construction, or operation is a backfit.
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