View source for JPN-97-012, Forwards Response to NRC RAI Re EDG TS Change

{{Adams
| number = ML20137A527
| issue date = 03/17/1997
| title = Forwards Response to NRC RAI Re EDG TS Change
| author name = Knubel J
| author affiliation = POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000333
| license number = 
| contact person = 
| document report number = JPN-97-012, JPN-97-12, NUDOCS 9703200361
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 7
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:123 Man Street e          WNte PWns New YOfx10601 014 681 6340 914 287.330'3 (FA)0 e   *
    #> NewYorkPower                                                                         t""vi"e"Lnt ee tv Authority                                                                            c*' "~ o"-
March 17,1997 JPN-97-012 United States Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, DC 20555

==Subject:==
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Response to NRC Request For AdditionalInformation Regarding Proposed TechnicaLSnec.ification Change _(JPTS-95-007)

==References:==
: 1. NYPA letter, W. J. Cahill, Jr. to USNRC Document Control Desk, (JPN-96-004), dated January 25,1996
: 2. NYPA letter, W. J. Cahill, Jr. to USNRC Document Control Desk, (JPN-96-018), dated April 26,1996
: 3. NYPA letter, W. J. Cahill, Jr. to USNRC Document Control Desk, (JPN-96-038), dated September 12,1996 4.-    USNRC letter, Karen Cotton to W. J. Cahill, Jr., dated February 4,1997

==Dear Sir:==

The Authority submitted a proposed Technical Specification (TS) Change Request (JPTS       007) for the James A. FitzPatrick Nuclear Power Plant (Reference 1). This TS proposes to revise allowed out-of-service times for single inoperable Emergency Diesel Generators (EDGs) to accommodate on-line maintenance of the EDGs. In References 2 and 3 the Authority responded to requests for additional information (RAls),
in Reference 4, the NRC provided an additional (RAI) dated February 4,1997. In the enclosure to this letter each question is restated, followed by the Authority's response.
9703200361 970317 PDR      ADOCK 05000333 P                        PDR 2000GG B155555!I1951.115

., , The Authority has reviewed References 1,2,3 and the enclosure to this letter and has concluded that this response does not change the proposed amendment or modify any of the conclusions stated in the Significant Hazards Consideration Evaluation previously submitted in Reference 1.
The Authority requests approval of the proposed amendment by April 14,1997 in anticipation of planned EDG maintenance scheduled during May 1997. Should you have any questions regarding this response, please contact Mr. Art Zaremba.
Very truly yours,
                                                                                             /
j
                                                            / James Knubel Senior Vice President &
Chief Nuclear Officer cc:    JPTS-95-007 File Regional Administrator                                                                     !
U.S. Nuclear Regulatory Commission 475 Allendale Road                                                                         3 King of Prussia, PA 19406                                                                  !
Office of the Resident inspector U.S. Nuclear Regulatory Commission                                                         1 P.O. Box 136                                                                               l Lycoming, NY 13093 Ms. Karen Cotton, Acting Project Manager                                                   l Project Directorate 1-1                                                                    l Division of Reactor Projects - 1/II                                                        I U.S. Nuclear Regulatory Commission                                                         1 Mail Stop 14 B2 Washington, DC 20555 Mr. F. William Valentio, President l
New York State Energy, Research and Development Authority                                  l Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399

    ,    .                                      Attachment to JPN-97-012
.             BeSp0Dse_10Reguest for Additional Information Related to Emergency Diesel Generator
;                                  Technical Soecification Change (TAC No. M94611)
Question 1:
A general description of the approved change to your work control process.                      !
Response 1:
Administrative Procedure, AP-10.02, "13-Week Rolling Schedule", provides the general description of our work control process. The following are the highlights of the process detailed in AP-10.02.
A required assessment of safety impact is performed for work with the unit on line.
A screening table lists maintenance rule systems with risk significance. The basis for risk significance is a combination of the PRA insights and the deterministic approach adopted by the maintenance rule expert panel by 4
considering the contribution to defense in depth of seven key safety functions:
i                        Electrical power distribution, containment integrity, containment heat removal, high pressure injection, low pressure injection, RPV pressure control, reactivity
;                        control. The Work Week Manager is required to review work activities to determine the potential effect on the ability for the system / component to
;                        perform the safety functions as listed for that system.
Once determined that a safety function is degraded, specific guidance is
;                        provided by a set of tables which contain various system / component configurations. This table identifies a level of risk for each configuration. The
 ;.                      table also includes guidance that has been developed to minimize the impact in j                        the event of transients during the specific maintenance configuration. If the configuration is not specifically analyzed, a dependence matrix and a set of logic trees are reviewed. The Work Week Manager determines if the configuration can be analyzed using this method. If the configuration can not be analyzed as stated above, the procedure dictates that the schedule be changed or the PRA group analyzes the configuration to determine the level of risk prior to commencing work.
Maintenance performed that has risk significance requires additional planning steps to reduce out of service time and to review potentially conflicting activities. Additionally, proportional to the level of risk, upper management approval is required up to and including the Site Executive Officer prior to commencing work.
Page 1 of 5

    ,   .                                     Attachment to JPN-97-012 Besoonse to Request for AdditiQual Information Related to Emergency Diesel Generator Technical Specification Change (TAC No. M94611) 4 Question 2:
i             A description on how risk information/ insights from your PRA is, or will be, incorporated
]               into your work control process, specifically with respect to planning and controlling
.;              maintenance activities involving the EDGs.
!         Response 2:
The response to Question 1 provided a general description of the work control processes which utilizes risk information and risk insights into plant maintenance decisions. In addition, with respect to on-line emergency diesel generator (EDG)
{               maintenance, the Authority previously described additional administrative controls iri j               Section Ill.a.3 of the safety evaluation enclosed with Reference 1. These controls ensure that the three remaining EDGs will be operable and available to mitigate the consequences of a LOOP condition and that the monthly operability test (ST-98) on the EDGs will have been performed within the previous two week period to support the i               on-line EDG maintenance.
]
4 m
  .                                                                                                        I i
l Page 2 of 5 1
1
                                                                                                           )

i 1
Attachment to JPN-97-012 Response to Request for Additional Information Related to Emergency Diesel Generator Technical Soecification Change (TAC No. M94611)
Question 3:
A discussion on how completeness is assured by consideration of equipment not                       ;
modeled in your PRA and how it is, or will be, addressed in your work control process.
Response 3:                                                                                                  !
As stated in the response to Question 1, AP-10.02 contains a set of tables which                    :
identifies a level of risk for various plant configurations. If a specific configuration is not listed in the set of tables, then the configuration is evaluated by use of a                    ,
dependence matrix and a set of logic trees. If this specific configuration cannot be                 1 analyzed, the procedure dictates that the work schedule is changed or the PRA group is consulted for further guidance prior to commencing work.
l l
l l
1 l
l I
l Page 3 of 5 1

     .                                     Attachment to JPN-97-012 Besp00se to Request for Additional Information Related to Emergen_cy Diesel Generator Technical Specification Change (TAC No. M94611)
Question 4:
A discussion on the extent to which numerical criteria, i.e., core damage frequency or incremental core damage probability, are, or will be, used in your work control processes.
Response 4:
Explicit numerical risk criteria, in the form of an incremental core damage probability, are used to evaluate unscheduled on-line maintenance. The criteria used are those proposed for temporary risk increases in EPRI TR-105396'.
The incremental core damage probability for the proposed on-line maintenance is calculated as the product of the increment of core damage frequency (CDF) for the temporary condition and the anticipated duration. Unscheduled on-line maintenance 4
with risk increases >10 is to be subject to further evaluation and management scrutiny. The Authority notes that to date no proposed unscheduled on-line maintenance at Fitzpatrick has had a temporary risk increase that exceeded 10 4 i
l l
l l
l l
l ERIN Engineering and Research,                         Inc.,     "PSA Applications Guide", Prepared for Electric Power Research Institute, EPRI-TR-105396, August 1995.
Page 4 of 5

,   .                                        Attachment to JPN-97-012 Besponse to Request for AdditionalInformation Related to Emergency Diesel Generator Technical Soecification Change (TAC No. M94611)
Question 5:
A list and brief description of your procedures that govem your risk-informed work control.
Response 5:
Administrative Procedure AP-10.02, "13-Week Rolling Schedule". This procedure defines the process of scheduling work activities based on a 13-week rolling or repetitive schedule. This procedure is intended to assess and control the effects on key plant safety functions of the removal of safety-related structures, systems or
                                                                                                           )
components (SSCs) from service while at power.                                              l Administrative Procedure AP-05.13, " Maintenance During LCOs". This procedure establishes controls for evaluation and managing risk associated with removal of SSCs from service to conduct work under a Umiting Condition of Operation (LCO). This procedure is intended to assess and control LCO maintenance by requiring planning steps to reduce out of service time, review potentially conflicting activities, and receive I upper management approval for maintenance on risk significant equipment.                    l 1
l l
l l
Page 5 of 5}}