ML20094P151
| ML20094P151 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/01/1992 |
| From: | Ralph Beedle POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JPN-92-018, JPN-92-18, NUDOCS 9204070255 | |
| Download: ML20094P151 (86) | |
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April 1,1992 JPN-92-018 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1 137 Washington, D. C. 20555
Subject:
James A. FitzPatrick Nuclear Power Plant Dot:ket No. 50-333 Station Blackout Rule Bgspgnso to Safety Evolu.;ttignEggg.ntntndationa
References:
- 1. NRC letter, B.C. McCabe to R.E. Beedle, dated November 13,1991, " Safety Evaluation of the James A.
FitzPatrick Nuclear Power Plant Response to the Station Blackout Rule" (TAC 66546).
- 2. NYPA letter, R.E. Beedie to the NRC, (JPN 91-066), dated December 18,1991, " Response to Safety Evaluation Recommendations."
Dear Sir:
Additional Authority responses to the recommendaticns of the NRC Safety Evaluation (Reference 1) era enclosed as Attachment 1. The Authority committed to provide this additionalinformation in Reference 2.
Full responses to three safety evaluation recommendations related to four hour station battery capacity calculetions, review of the station blackout procedure, and implementation of an EDG reliability program, will be provided by Decemoer 21,1992.
The specific criteria used to exclude or include each containment iscia' on valve from station blackout operational and position indication guidelines are identified in Attachme.its II, Ill, and IV.
If you have any questions, please contact J. A. Gray, Jr.
Very truly vours, l
r D-(b Ralph E. Beedle Executive Vice President Nuclear Generation cc: Next Page 060 9204070255 420401 g'
PDR ADOCK 05000333 l
p PDR
JPN 92 018 April 1,1992 Page 2 cc:
Regional Administrator U.S. Nuclear Regulatory Commission
- 475 Allendale Road King of Prussia, PA 19406 Office of the Resident inspector U.S. Nuclear Regulatory Commission P.C. Box 136 Lycoming, NY 13093 Mr. Brian C. McCabe Project Directorate I 1 Division of Reactor Projects -l/il U.S. Nuclear Regulatory Commission Mail Stop 14 82 Washington, D.C. 20555 Attachments:
1 Additional Responses to Safety Evaluation Recommandations il Containment isolation Provisions During an SBO - Valves Concern lli Containment isolation Provisior's During an SBO - Excluded Penetrations IV Containment isolation Provisions During a Station Blackout -
Individual Penetration Data Sheets t-i
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Attachment I to JPN - 92 018 April 1,1992 11 2
J STATION BLACKOUT RULE Additional Responses to Safety Evaluation Recommendations i
!l' New York Power Authority James A. FitrPatrick Nuclear Power Plant Docket Number 50-333
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Attachment I to JPN-92-010 Station Blackout Safety Evaluation Additional Responses - April 1,1992 Intrash!ction This attachment provides additional responses to the NRC Safety Evaluation (Reference 1 *)
conceming station blackout. The Authority provided an initial response in Reference 2'.
The Authority's initial responso did not provide complete answers to all of the safety ovaluation recommendations. For those items which were not fully addressed, the Authority committed to provide more compl9te answers, or completion dates for those answers. This attachment provides those additional responses and submittat dates.
Prior to receipt of the safety evaluation, the Authority had responded to NRC requests for additionalinformation (Reference 3'). That Authority response was received too late to be considered in the NRC safety evaluation. The response did, however, address many of the concerns subsequently presented in the NRC safety evaluation. Accordingly, this response occasionally refers to the answers previcusly provided in Reference 3.
The safety evaluation makes sevemt recommendations concerning operability and position indication for electrically interlocked containment isolation valves (CIV) that the Authority exchvjed from SBO procedures, in addition to the response contained in this attachment, the Authority has provided Attachments 11, til, and IV to identify, for cach CIV, the basis for excluding or including the valve or penettstion as an area of SBO concern.
The NRC safety evaluation recommendations are repeated in italics. The Authonty's
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response follows the NRC recommendation.
- Note: References are identified on page 14 of this attachment.
Page 1
Attachment i to JPN 92-018 Station B!ackout Safety Evaluation Additional Responses - April 1,1992 Remanma11]ndlyldmaLanfity_EyJ!Wu!QaltG9mmtDdall0H3 i
NRC Recom.Lnendation:
2.2.2 Class IE Battery Capacity The licensee should verify and confirm that, with the load shedding proposed for SBO conditions, the batteries have sufficient capacity, plus aging and load growth margins to cope with and recover from an SBO of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The verification of battery adequacy should consider loads such as those that are needed for EDG start attempts and switching requirements (breaker controls) at the beginning and ths end of the S80 event.
NYPA Resognggi Details of the load shedding scheme and calculations to demonstrate the ability of the station batteries to power SBO loads for four hours were described in Reference 3. This calculation was conducted in accordance with the procedures in effect in 1989. The SBO battery capacity calculation (JAF 89-013, revision 2) considered EDG field flashing for EDG start attempts and circuit breaker controlloads for breaker reclosure during the load shedding and reconnection sequences for restoration of ac power as loadt at the onset of the event. These two loads were not considered as random loads throughout the event nor were they considered as loads during the last minute of the four hour coping period. However, this calculation indicated that for the limiting interval at the end of the duty cycle, the sixteen r.ositive plates in each battery exceeded the calculated minimum numbe of roquired plates (12 for 72SB-2 and 13 for 72SB 1) by a significant margin. This calculateo margin was considered to be adequate to include EDG field flashing and breaker reclosure loads even curing the last minute of -
the four hour coping period.
A new formal engineering review is being conducted in accordance with current engineering design procedures. This review started with the calculations supporting the two hour duty cycle described in the FSAR. This review encompasses the updating, revision and unification of several existing battery capacity calculations including those performed for the original plant l
. design basis, station battery replacement, and the SBO battery capacity calculation addressed in Reference 3.
Upon completion of the review for the FSAR two hour duty cycle calculations, the calculations will be revised to address SBO criteria and assess the battery capacity relative to the SBO four hour coping period. This adjustment is necessary to er'commodate the differences in the battery loads and the duty cycle time limits between the SBO and FSAR criteria.
The Auth'arity anticipates that the revised calculations will be completed by December 21,1992.
Page 2
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Attachment I to JPN 92 018 Station Blackout Safety Evaluation -
Additional Responses - April 1,1992
' MRC Recommendation:
i 2.2.5 Containment isolation Add the valves that were excluded by the additional criteria (a valve interlocked with another valve) in an appropriate procedure.
Identify actions which are needed to confirin these valves (interlocked valves excluded by methodology) are closed. The valve closure needs to be confirmed by position indication (local, remote, mechanical, process information, etc.) independent of the preferred (off site) or on site power.
NYPA Response:
The Authority has identified 15 valves of concern for the isolation of six containment penetrations during an SBO event. All of these valves are operated by de motors and have position indication in the control room.
These valves and penetrations are identified in Attachroent II.
The Authority has not added the valves excluded due to the interlocked criteria, into plant procedures. The reasons for this are crosented in tho discuss;on of the penetration-by-penetration methodology, and explanations of the criteria used for excluding interlocked valves and water sealed valves, which are presented in this attachment.
Accordingly, the Authority does not consider it nvcessary to confirm that interlocked or water sealed valves (excluded by methodology) are closed during an SBO. The Authority does not consider position indication to bo needed during an SBO, for the valves excluded by these two criteria. This position is supported by the detailed descriptions of the 26 valves oxcluded
' by these criteria provided below.
Eight valves isolating four penetrations were excluded using the series mounted electrically interlocked valve criteria. Eighteen more valves isolating
. seven penetrations were excluded using water sealisolation criteria. These additional exclusion criteria, and the application of the criteria to the 26 valves excluded, are discussed in detaillater in this attachment. Most of the remaining valves and penetrations have been excluded from concern during _
The valves and penetrations which were excluded from SBO concern are identified in Attachment 111. This attachment identifies the specific NUMARC or other criteria, together with the SBO evaluation, that were used to determine the exclusion for each individual valve and penetration from concern during an SBO.
/Jtachment Ili supersedes and replaces " Table 8" which was previously provide M the NRC in Reference 3.
Some l_
information contained in " Table
- mrning the availability of position i.
indication on the 27 MAP panel was not conet.
Page 3
Attachment I to JPN-92-018 Station Blackout Safety Evaluation Additional Responses April 1,1992 NYPA Resoonso (continuedh Sketches of each penetration showing the location and arrangement of the excluded isolation valves, together with a short table of data for each valve are provided as Attachment IV. These sketches are provided only as an aid to understanding the application of the interlocked and water sealed valve criteria to specific installations. The sketches and the associated tabular data have not been reviewed, approved or controlled in accordance with procedures for formal engineering calculations and drawings.
Methodology: NUMARC criteria uses a valve-by-valve approach to identifying every valve on every containment isolation line and verification of closure. One of the problems with this criteria is that the shorter the coping period determined in accordance with 10 CFR 50.63, the less time is available to close or check every valve. At the FitzPatrick plant, the relatively short required coping period of four hours makes it difficult, if not impossible, to manually close or check the position indication for e4 OlV valve before the coping period for the SBO event would have endeo ^ >r example, some of the CIVs are not accessible without the use of scaffolding.
In contrast to the valve-by valve approach, the Authority used a penetration-by-penetration verification. At least one valve on every penetration meets one of the fivo NUMARC criteria, or the interlocked valve or water seal criteria. Some penetrations such as TIP and feedwater have been excluded for other reasons identified in Attachment 111. For example, feedwater check valves (which do not have control room position indication) have been accepted in accordance with Dagulatory Guide 1.97 which excluded check valves from requirements for position indication. Valves acceptable under RG 1.97 for LOCA conditions should also be acceptablo for an SBO event.
In the Fit:rPatrick methodology, if two Valves were in series and one valve l
met an exclusion criterion, the penetration was excluded even if the second valve did not meet the critoria. This approach provides adequate assurance that all containment penetrations are isolated by at least one valve during an SBO. This approach is consistent with Reference 4 which states that it is not necessary to show that both the inboard and outboard containment I
iso!ation valves can be closed during an SBO.
Interlocked Valve Exclusion Criterion: A large number of containment isolation valves are closed during reactor operation in many cases, the closure of these valves is ensured through interlocks that do not permit the -
penetration to be opened under operating conditions. This category includes l
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CIVs that isolate low pressure systems from high pressure reactor coolant piping. These valves have been excluded from consideration if interlocks ensure closure of the penetration by at least one CIV during normal l
operation. With one valve closed and the other valve open on the samo l
penetration, the interlock assures that the closed valve remains closed so long as the other valve is open. An example is the core spray injection line which is isolated by two ac motor operated valves installed in series.
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Attachment I to JPN 92-018 Station Blackout Safety Evaluation Additional Responsas - April 1,1992 NYPA Resoonse (continug.dk Although one ClV may be open during normal operation, interlocks do not permit the second CIV to open unless the reactor pressure is less than 450 psig. (Reactor operating pressure is normally 1000 psig.)
The table previously submitted in Reference 3 identified penetrations with interlocked CIV3 installed in series which isolate the same line. This criteria does not apply if the valves were installed on different lines.
The NUMARC criteria would require closing the second valve even though the first valve was already closed. During an SBO, ac power would not be available to close the second valve. To meet the criteria, the open valve would have to be closed manually. Manually closing all of the redundant or interlocked valves, even assuming that a sufficient number of licensed operators could be assembled and devoted exclusively to this task, would not be an efficient use of personnel, does not significantly improve the level of protection provided to the public, and would be likely to require more time to impleinent than the four hour SBO coping duration.
Application - Exclusion of eight valves using the interlocked valve criterion:
The series mounted electrically interlocked valve criterion has been used to exclude only four penetrations.
Penetrations 39A,B are for the drywell spray lines supplied from the residual heat removal (RHR) system. Isolation of these lines is provided by ac motor operated valves 10MOV 31 A,B and 10MOV 26A,B. Therefore, these valves would become inoperable during an SBO event. However, each pair of valves (10MOV-39A and 10MOV-26A) and (10MOV-398 and 10MOV-268) are electrically interlocked and mounted in series. The interlocks insure that one valve is always closed except during drywell spray operations. Drywell spray requires operation of a key lock switch in the control room and a high drywell pressure signal to override this interlock.
Penetrations 211 A,8 are for the torus sprw line supplied from the residual heat removal (RHR) system. Two ac pows id CIVs (10MOV-38A,B and 10MOV-39A,B) are installed in series on each torus spray line. These two CIV valves are electrically interlocked so that when one valve is open, the other must be closed. Thus, one valve is always closed and the line is isolated during routine plant operations. The interlock can only be overridden by use of a key lock svvitch in the control room and a high containment pressure signal.
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Attachment I to JPN 92-018 Station Blackout Safety Evaluation Additional Responses Apil 1,1992 NYPA Resoonse_1gontinued):
Application - Exclusion of eighteen valves using the water seal criterion:
An example is the suppression pool cooling line, which taps into the torus
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spray line between the two valves (10MOV-38 and 10MOV-39) and terminates at the torus (panetration 210A,B) below the minimum suppression pool water level. These torus cooling lines have one isolation valve (10MOV-34A,B) installed on each iine. A sketch of this arrangement is included in Attachment IV.
During torus cooling operations, the inboard isolation valves 10MOV-38A,B would be closed isolating penetrations 211 A,B. The outboard isolation valves 10MOV-39A,B would be open along with the suppression pool cooling isolation valves 10MOV-34A,B. This would appear to create an unisolated potential pathway for leakage out of the primary containment during an SBO when power would not be availsble to operate 10MOV-34A,B and 10MOV-39A,0.
However, because the torus cooling line opens within the torus below the minimum suppression pool water level, a water seali:, created. The water seal effectively insures isolation of penetration 210A,8 during an SBO even
= though the two isolation valves would not be not electrically operable.
Becaur.e the lino opens into the torus below the minimum suppression pool water level, the line remains full of water at all times. Although part of this water sealis outside the primary containment, the NUMARC guide lines assume that no other failure will occur simultaneously with an SBO.
- Therefore the integrity of the water sealis assumed to remain intact during an SBO. Accordingly penetration 210A,B and valve 10MOV-34A,8 have been excluded based on the water seal created by the piping extending below the minimum suppression pool water level.
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. Attachment I to JPN-92-018 Station Blackout Safety Evaluation Additional Responses - April 1,1992 NRC Recommendallom 2.2.5 Containment isolation (continued)
In addition, the licensee should verify that the fait closure of the drywell and torus pressure sensing penetration valves would not cause the loss of pretsure indications in the control room for these areas.
NYPA Resoonse:
The instrumentation which would be available to the operator during an SBO and loss of uninterruptible power supply (UPS), was identified in Reference 3.
The cor.cern expressed in the recommendation is believed to be related to the function of two penetrations which are identified in FSAR Table 7.3-1 as follows:
Containment Penetration Valve isolation Close Normal Penetration Function Number
$10DAL_
li.ma Status 45 Drywell 161 AOV 101 A A,F,R,Z N/A Open Pressure Sensing 161 AOV-101 B A.F,R,Z N/A Open 218 Torus 16-1 AOV-102A A,F.R,2 N/A Open Pressure Sensing 16-1 AOV 1028 A,F,R,Z N/A Open These two pressure sensing lines are used only for the integrated leak rate testing (ILRT) instrumentation. They do not provide information to the operator during normal operation nor during an SBO. Therefore, isolation of these lines during a station blackout event is c:ceptable. Control room primary containment pressure instrumentation is served by penetration number 50c " Instrumentation Sensing DW Pressure". Penetration 50c is shown as " typical" and the listed characteristics of the penetration also apply to the instrumentation line penetrations for control room indications of containment pressure.
As a part of the July 1992 update of the FSAR, the Authority will add notations to FSAR Table 7.3-1 in the " Remarks" column to indicate that these two lines are used only for ILRT testing.
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1 Attachment I to JPN 92-018 o
Station Blackout Safety Evaluation Additional Responsos - April 1,1992 NRC Commitment 2.3 Procedures and Training The licensee stated that plant procedures have been reviewed and changes necessary to meet NUMARC 87-00 willbe implementedin the following areas:
- 1. Station Blackout response (Procedure Nc. AOP-49, " Station Blackout' will conform to NUkfARC 87-00, section 4.2. t.
NYPA Resoonse:
The procedure will be reviewed and compared to NUMARC 87-00. The changes necessary to meet the guidance of NUMARC 87-00, section 4.2.1 will be implemented by December 21,1992.
NRC Excectation:
2.3 Procedures and Training The staff expects the licensee to implement the appropriate training to ensure an effective response to an SBO.
- NYPA Resconse:
The 1991 fourth quarter licensed operator reaualification simulator training was conducted in accordance with training department coursollesson plan R91-7.1.-The licensed operator cycle 7 simulator training was conducted between 9/23i91 and 10/31/91. The final scenario consisted of a walk through of a station blackout sequence.- The scenario was frozen approximately 20 seconds after its initiation. A panel walk down and discussion was then conducted in order to emphasize equipment status, plant parameter indication status, and desired course of action. Following the discussion, the simulator was shifted to "run" to allow the crew to perform the actions of AOP-49, " Station Blackout," up to the steps for restoring electrical busses. The remainder of this scenario _was considered to be similar enough to the restoration _ from a previously simulated event (loss of power due to failure of transformer T1 A ) to justify the omission of the restoration of power phase of the SBO event.
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Attachment I to JPN 92-018 Station Blackout Safety Evaluation Additional Responses - April 1.1992 o
NYPA Resoonse (continuedli The station blackout scenario will be reviewed periodically in requalification training once every two to four years.
A clessroom lesson plan is scheduled for development during 1992 for use with the current license replacement class, it will be developed and taught before October 1992 and will become part of the licensed operata replacement training program permanent curriculum.
NRC Recommendation; 2.4 The licensee should include a full description, including the nature and objectives of the required modifications identified above in the
-documentation that is to be maintained by the licensee in support of the 380 submittals.
Note: " identified above
- refers to two modifications. One modification will provide an alternate power source to the RCIC enclosure ventilation fans which will eliminate the potential for RCIC system isolation on high ambient temperature during an SBO. The other modification will provide alternate power to selected instrumentation, under SBO conditions, to provide operators with information that would otherwise be lost upon shedding of the UPS MG set after one hour into the SBO.
NYPA ResoAn.1qi Reference 2 stated that "The design of both modifications is expected to be finalized in February 1992." and that "A more complete description including the nature and objectives of the modifications will be provided by March 23, 1992."
Descriptions of the nature and objectives of *he proposed modifications are provided below. These descriptions represer. the proposed designs currently in the review and approval process. Changes may be mado during this process. The Authority will provide revised descriptions to the NRC if significant changes are made to the nature or objectives of these proposed modifications. Detailed engineering descriptions of both modifications will be completed three months prior to the outage in which they are to be installed. Both modifications are scheduled to be installed during the Reload 11l Cycle 12 refueling outage currently scheduled to begin in October,1993.
The modification documentation maintained in plant records willinclude a fuli description including the nature and objectives of the modifications.
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Attachment I to JPN 92 018 Station Blackout Safety Evaluation Additional Responses - April 1,1992 -
NYPA Resoonse fcontinuedh The Authority will provide the NRC with an explanation and justification for a delay in completion of the modifications if changes to the outage starting date would result in installation of the modifications beyond the two year limitation of 10 CFR 50.63(c)(4) which expires on November 18,1993.
RCIC Enclosure Ventilation Fan Power Supply: (F189-159) The two fans which provide ventilation to remove heat from the enclosure for the RCIC system are currently powered from two senarate motar control centers (MCC). Fan 13FN-2A starts if the temperature in the ?:C enclosure rises to 90*F. The MCC for this fan receives power from normal station ac power.
Any loss of normal station ac power would rendor this fan inoperable.
Fan 13FN-1 A starts if the temperature in the RCIC enclosure rises to 105'F.
The MCC supply to this fan,in addition to receiving normal station ac power, is backed up during loss of normal station ac power by reserve station power from off site and from the on site emergency diesel generators. However, during a postulated SBO event, power would also be lost to this fan and it would be inoperable.
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Therefore, during a station blackout, the power to both ventilation fans would be lost and the temperature of the RCIC er' closure would increase.
The increased temperature could potentially impair the operation of some component; of the RCIC system.
To provide for operation during an SBO, a backup power supply to both fans l
will be added from the OA category I Low Pressure Coolant injection (LPCI) division I dc to ac inverter. Upon loss of all ac power, the inverter will convert de power, from the division i LPCI battery to ac power to supply the RCIC enclosure ventilation fans. Both of the RCIC ventilation fans will bo powered from this LPCI battery during an SBO. The LPCI battery and inverter have adequate capacity to power these fans and other required equipment during an SBO. The transfer of the power source to the LPCI inverter will be automatic without operator action. Upon completion of the modification the fans will be reclassified as OA category M.
The Authority previously stated in Reference 3 that this modification was y
tentatively scheduled for installation during the 1992 refueling outage. The j
schedule for installation of this modification has been changed to the Reload _11/
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Cycle 12 refueling outage currently scheduled to start in October,1993.
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i Attachment I to JPN-92-018 Station Blackout Safety Evaluation Additional Responses April 1,1992 1
i NYPA Resoonse (continued):
Monitoring and Analysis Panel 27 (27 MAP) power supply: (F1-89158) The panelis currently powered only from the station ac power. The modification will provide automatic transfer to an alternate dc power source upon loss of all ac power sources (SBOL The modification, as previously described to the NRC (Reference 3), would have provided alternato power from the LPCI inverters which in turn are supplied by the LPCI batteries. The proposed source of dc power is currently being reviewed and may be changed to the station battery.
Alternate ac power will be provided from new inverters to be installed in the vicinity of the 27 MAP panel. When station ac power is lost (SBO), an under-voltage relay willinitiate the automatic transfer from the ac supply to the de power supply which will ener0 ze the 27 MAP panel ac loads from the new i
local inverters. The inverters will be classified as OA category 1.
The 27 MAP panel also supplies a number of de loads. Currently, these loads receive power from ac-to-dc rectifier power supplies. During an SBO event, the ac power to the rectifiers would be lost and consequently the de loads would be lost. This modification will install additional de backup modules inside the 27 MAP panel. During an SBO event, the new de backup modules will automatically connect each de load to a source of de power (station battery or LPCI battery).
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Attachment I to JPN 92 018 Station Blackout Safety Evaluation Additional Responses April 1, '992 NRC Excectation:
2.5 Quality Assurance and Technical Specifications The staff expects that the plant procedures willreflect the appropriate testing and surveillance requirements to ensure the operability of the necessary SBO equipment.
The licensee shou ld verify that the SBO equipment is covered by an appropriate QA orogram consistent with the guidance of RG 1.155. This evaluation should be documented as part of the documentation supporting
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the SBO rule response.
NYPA Resoonsel I
SBO equipment coverage by QA program: The plant equipment and systems necassary to meet the requirements of 10 CFR 50.63 (Station Blackout) are corrently classified as quality assurance (OA) category 1, safety related.
Equipment classified as OA category 1 is controlled under the site Nuclear Safety Related Quality Assurance Program which was established to meet the requirements of 10 CFR 50, Appendix 0. This program is described in Chapter 17 of the Final Safsty Analysis Report. Implementation of the program is assured by administrative procedure AP 1.7,"Ouality Assurance Program implementation at JAF". This program applies to equipment and activities classified as either QA category I or category M.
NRC Regulatory Guide (RG) 1.155, " Station Blackout" in section C,
" Regulatory Position", paragraph 3.5 " Quality Assurance and Specification Guidance for Station Blackout Equipment That is Not Safety Related", states in part that:
" Appendices A and B provide guidance on quality assurance (QA) activities and specifications respectively for non safety related equipment usad to meet the requirements of 10 CFR 50.63 and not already covered by existing QA requirements in Appendix B or R of Part 50."
The quality assurance program elements identified in Appendix A of RG 1.155 are encompassed entirely within the QA program elements of the FitzPatrick Nuclear Safety Related Quality Assurance program which meets the reauirements of 10 CFR 50, Appendix 8.
The equipment required to meet station blackout requirements is classified as OA category 1. QA category I equipment is controlled under the provisions of -
the site QA program. The OA program encompasses the guidance of Appendix A to RG 1.155. Therefore, station blackout equipment is covered under an appropriate QA program consistent with the guidance of RG 1.155.
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Attachment I to JPN 92-018 Station Blackout Safety Evaluation
- Additional Responsos - April 1,1992 NYPA Resoonse (continuedh Surveillance testing: Equipment and systems which are classified as OA category I are subject to the inspection and testing requirements of the OA program. Because the equipment necessary to meet the SBO rule is classified as OA category 1, the existing plant surveillance testing procedures reflect requirements which are appropriate to ensure the operability of the SBO equipment.
Use of equipment not classified as QA category 1: Although compliance with 10 CFR 50.63 requirements can be achieved using only equipment which is classified as OA category 1, the preferrod method of coping with a postulated station blackout involves the use of some equipment which is not classified as category 1.
As detailed in Abnormal Operating Procedure AOP 49 " Station Blackout",it is preferable to run the QA Category M Reactor Core Isolation Cooling (RCIC) system instead of the QA Category i High Pressure Coolant injection (HPCI) system during a blackout. The flow rate of the RCIC system more closely matches the SBO makeup requirements. The use of RCIC also minimizes the
_ potential for level 8 high reactor water isolations, which in turn minimizes the potential for unnecessary cycling of de motor operator valves and, therefore, extends battery power availability.
Although the preferred source of water for both HPCI and RCIC is from the condensate storage tanks (CSTs), water from the suppressien pool (torus) can be used. The torus is classified as a OA category I structure. Section 3.1.3.3 of the Individual Plant Evaluation (IPE) documents the maximum allowable cooling water temperature for sustained HPCI or RCIC turbine operation as 200 'F.
Since the predicted time to heat up the torus to 200 *F -
is in excess of the SBO determined four hour coping duration, the_ torus would be sin acceptable source of water for either HPCI or RCIC.
1 Page 13
Attachment I to JPN 92 018
-1 Station Blackout Safety Evaluation Additional Responses April 1,1992 1
NRC Recommendatiorn 2,6 EDG Reliability Program It is the staff's position that an EDG reliability program should be developed in accordance with the guidance of RG 1.155, section 1.2. If an EDG reliability program currently exists, the program should be evaluated and adjustedin accordance with RG 1.155, Confirmation that such a program is in place or will be implemented should be included in the documentation that is to be maintained by the licensee in support of the SBO submittals.
NYPA Resoonit-EDG performance has consistently surpassed, by a significant margin, the reliability goals outlined in Appendix F to NUMARC 87 00.. The EDG reliability for the past 20,50, and 100 demands was presented in Reference 3.
The Authority willimplement a program incorporating the guidance contained in Regulatory Guide 1.155 by December 21,1992. The Authority may revise this program when the NRC's unresolved safety issue B-56
" Emergency Diesel Generator Reliability" is resolved.
REFERENCEL 1.
NRC letter, B.C. McCabe to R.E. Beedle, dated November 13,1991,
" Safety Evaluation of the James A FitzPatrick Nuclear Power Plant Response to the Station Blackout Rule (TAC 68546)."
2.
NYPA letter, R. E. Beedle to the NRC, (JPN 91-066) dated December 18,1991, " Response to Safety Evaluation Recommendations."
3.
NYPA letter, R. E. Beedle to the NRC, (JPN-91-049) dated September 13,1991, " Response to Request for Additional Information Regardira Station Blackout."
4.
" Guidelines and Technical Bases for NUMARC initiatives Addressing Station Blackout at Light Water Reactors", August 1991, NUMARC 87-00 Rev.1, Appendix I " Responses to Questions Raised at the NUMARC 87-00 Seminars", page 122, question 101.
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_y. 1 to JPN 018 April 1,1992 STATION BLACKOUT RULE l
Containment isolation Provisions During an SBO Valves of Concern I
(Necessary for Isolation During an SBO) i l
l.
New York Power Authority James A. FitzPatrick Nuclear Power Plant Docket Number 50-333 i
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Attachment lll to JPN 018 Apto 1,1992 STATION BLACKOUT RULE Containment isolation Provisions During an GBO Excluded Penetrations l
New York Power Authority James A. FitzPatrick Nuclear Power Plant Docket Number 50-333
Attachment III to JPN-92-018 James A. FitzPatrick Nuclear Power P! ant Containment isolation Provisions During Station Blackout EXCLUDED PENETRATIONS Nominal NUMARC 87-Penetration Valve 00 Number Size (in.)
System Valve fDs Exclusion SBO Evaluation Cntena 7A B.C.D 24 Main Steam 29AOV-Note 1 NUMARC 87-00 Assumption 2.4.1(2) aHows utshtres to 80A(B.C.D1 assume proper operation of the MStVs. MSIV isolatsen j
would occur automaticatty on
- Low Condenser Vacuum
- 29AOV-Nrte 1 o, by operator action dunng a SBO.
86 A(B,C,D) s 9A 18 Feedwater. RCIC, 34FWS-28A 3
Penetration isolate ensured by check valve 34AVS-28A.
13MOV-21 12MOV-69 SB 18 Feedwater & HPCI 34FWS-28B 3
Penetration isolation ensured by check valve 34FWS-258.
34NRV-1118 3
23MOV-19 13A,8 24 RHR (LPCI) 10AOV-68AtB) 3 Penetration isolaton ensured by testable check valve 10AOV-68A{B).
10MOV-25A(B) 10MOV-27 A(E)
I 16A,8 10 Core Spray 14 AOV-13 A(0) 3 PeWrate isolation ensured by testable check valve 14AOV 13A(B).
(
14MOV-11 A(B)
I 18 3
20AOV-83 fads clored on loss of a r o-AC and ensures I
Sump Discharge penetration isolation-l 20MO9 A t
t 19 3
20AOV-35 fads c:ostd on less of air ci AC and ensures.
{
Drain Sump penetraten esofation.
D:scharge 20MOV 94 i
1 Page 1
Attachmeat lit to JPN-92-018 James A FitzPatrick Nuc! ear Power Plant Containtrent isolation Provisions During Station B!schout EXCLUDED PENETRATIONS I
Nominal NUMARC 87-Penetration Valve 00 Number Size (in.)
System Vafve ids Exclusian SBO Evabation Cntere b
22 1
N, Supply to DW 29tAS 22 3, 5 Penetration excluded based on 1" rw)mtral sire.
Penetraten isolation also ensured by check vatve 391AS-27SOV-141 5
22-23,24 4
RBCLCW to DW 15AOV-130A (B)
Penetration isolaten ensured by check vafves 4SESW-16B(A) and 15RBC-24A(P).
46ESW-16B(A) 3 1 bR8C-24 A(B) 3 25,71 18 DW Purge inlet 27 AOV-112 2
Penetration isolation ensured by AOVs failers dosed on (Air) 5ss of air and check valves 27 CAD 48 and 27 CAD 69.
18 27 AOV-111 2
14 27AOV-131 A 2
(
14 27 CAD 46 3
14 2 7 AOV-131 B 2
t 14 27 CAD 49 3
6 26A 24 DW PurDe inlet 27 AOV-113 3
24* penetration isolation ensured by AOVs fa&g closed (Air of N,i on toss of air. Bypass valves and pipewithirmpe SOVs i
24 27AOV-114 3
are less than 3 nominal size.
2 27MOV-113 5
2 27MOV-122 5
i 3/8 27SOV-119F1 5
3/8 27SOV 119F2 5
3/8 27SOV-12CE1 5
3/8 2750V-120E2 5
f 3/8 27SOV-122E1 5
[
3/8 27SOV-122E2 5
l Page 2 I
r Attahment III to JPN-92-018 James A. FitzPatrick Nuclear Power Plant Containment isolation Provisions During Station Blackout EXCl.UDED PENETRATIONG 4
Nominal NUMARC 87 Penetration Valve 00 Number Site On.)
System Valve ids Exclusion SBO Evaluaton Criteria 31 Ac,Bc 1
RWR Pump Seal 02-2RWR-13AiB)
- 3. 5 Penetration exctoded based on 1 nominal sire.
Purge Additiona!!y. isolaten ensured by check valve 02-2RWR-02-2SOV-OO1 5
13A(S),
02-2SOV402 5
02-RWR40A(B) 5 i
31 Ad.Bd 1
DW Atmosphere 27SOV-135C 5
Penetration exctoded based on 1* nomrnal sire.
4 Suct+on 27SOV-135A 5
2750V 135D 5
2750V-1358 5
35A-D 1.5 Tips A through D 07SOV-104 A 5
Penetration excluded based on 1.5 nomme size.
07EV104A 5
)
[
0730V 104C 5
37A D 1
Cor. trol Rod Drive SOV-123 5
Penetrat: ens excluded based on 1" norranal size.
(Inlet)
SOV-120 5
AOV-126 5
CRD-138 3, 5 38AO 1
Control Rod Drive SOV-122 5
Penetratsons excluded based on 1* nominal size.
! Outlet) i SOV-121 5
i AOV-127 5
i Page 3 m.
. +
~
w
Attachment lit to JPN-92-010 James A. FitzPatrick Nuclear Power Plant Containment isolation Pro,isions During Station Blackout L
EXCLUDED PENETRATIONS
[
Nominal NUMARC 87-i Penetration ti Valve 00 Number j Site fin.)
System Valve ids Exclusion SBO Evaluaton
~
Critet.;
39 A..B 10 RHR DW Spray 10MOV-31 A(B)
Note 2 Penetration excluded based on electncal interlock preventog the opening of 10MOV-31 A(B) and 10MOV-10MOV-26 A(B)
Note 2 26A(B) under normat operating condit.ons.
[
41 1
RWR Lcop Sample 02-2SOV-39
- 2. 5 Penetratx>n excluded based on 1* nomenal sire.
02 2SOV-40
- 2. 5 42 1.5 SBLCS 11 SLC-17 3, 5 Penetrat.on excluded based on 1.5* nom nal sire.
Additkeaffy, isolation ensured by check va*ves 11SLC-1/
11 SLC-16 3, 5 and 11SLC-16.
11 EV-14 A(B) 5 45 0.5 DW Pressure 16-1 AOV-101 A 5
Penetratx>n excfoded base on 0.5" nomenal sire Ttws Sensing i
penetrat on is pressure sensing for ILRT instrumentation 16-1 AOV-101B 5
50c O.75 Instrumentation various 5
Penetraton excluded based on 0.75* nominal site.
Sensing DW Pressure 52a.55b 1
DW Atmosphere 27SOV-125C 2, 5 Penetrations excluded cased on 1* nommat site.
2?SOV-125 A 2, 5 s
- 2. 5 r
- 2. 5
~7c 1
Penetratens excluded based on 1* nomenal site.
DW L
Instrumentation 391AS-29 3, S 391AS-28 f
i Page 4 I
t
~
a.
n Attachment Ill to JPN-92-018 James A. FitzPatrick Nuclear Power Plant Containrnent isolation Provisions During Station Blackout EXCLUDED PENETRATIONS Nominal NVMARC 87-Penetrotson Voive 00 Number Site On )
System Vefwe ids Exclusion SB0 Evolustion Cnteria 58b.c 3/8 DW Hydrogen 2750V.122F2 2, 5 Penettet ons anchused based or. 3/8* non=nel aire.
54mple 2750V-122F1
- 2. 5 27S0 A120F2
- 2. 5 27SOV-120F1
- 2..,
58d,59 3/8 OW Hydrogen 2750V-123F2 2, 5 Penetretwns e=clueed beced on 3.9* remsnel este S ampia 2750V 123F1 2, 5 2750V 123E2 2, 5 2750V 123E1
- 2. 5 62,66 4
RBCLCSTSW Return 15 AOV-131 AtB) 4 Penettstmns enciuJed based on ckweed soop morde DW not from " A. B' DW evpected to be bmached in en 580-Cool.ng Assembly 15RBC-26 A18)
A 63.67 4
RBCLCSTSW to *B" 15 AOV-132 A(B) 4 Penetrat+ons e=ciudad based on closed bop moede DW not er Return from " A" espected to be breeched en en SBO.
15RBC-21 AtB)
- 3. 4 f
RWR Pump and Motor Coofere i
- 3. 4 l
Fenetratmne endeded based on closed loop inside DW not l
64,EB 4
RBCLCS/ESW Return 15AOV 133 AIBl 4
l from " A. B" RWR e@ected to be beesched donng en SSO.
Punw eM Wtor 15RBC-22 A;B) 4 Cociers 65 1.5 RBCLCS ESW Rerum 15 AOV 134A 5
Peretraten escheded bssed on 1,5* refW site.
from DW Equ$ ment Sump Coorer 159BC 33 5
Penetteten esolation ermred by check valves 27VB 6 and 27VB-2023 20 Reactor Buildmg to 27 AOV-101 A 1orus Vecuum 7.
Break ers 27VR6 3
27AOV 10tB 27 VB-7 3
Page 5 l
(
Attachment fit to JPN-92-018 James A. FitrPatrick Nuclear Power Plant Containment isolation Provisions During Station B!ackout EXCLUDED PENETTATIONS Nominal NUMARC 87-Penetration Va!ve 00
(
Number Size On.)
System Valve ids Exclusion SBO Evaluation Criteria l
l 202E 1.5 RCIC Turbine 13RC?C-11 5
Penetration excluded basal ori 1.5* nomina: size.
Exhaust tire Vacuum Breaker 13RCIC-37 5
' 3 MOV-130 5
27VB-5 3
203A 3/8 Sirporession Poot 2750V-119E2 5
Penetraton excluded based on 3/8" nominal size.
Atmosphere Sampie Suction 27SOV-119E1 5
2038 1
Primary 2750V 124E2 5
Penetration excluded based on 1" riomenai sire.
Containment 2750V-124E1 5
Analyrer and Post.
Accident Sample 2750V-124F2 5
1 Return 27SOV 124F1 5
205 20 Suppression Poot 27 AOV-117 2
Penetration excluded based on AOVs faibog closed on loss of air or AC. Bypass hne excluded tased on 2* nom nal Purge Exhaust 20 27 AOV-118 2
s,,,,
2 27MOV-117 5
2 27MOV-123 5
210A 16 RHR SPC Test 10MOV-34 A Note 3 Penetration exct.>ded based on a loop seal provided by a menemum suppression pool levei Return Line 4
RHR Pump 10MOV-16 A Mentmu" Tiow Une 4
RHR Heat 10MOV-21 A Exchanger Drain Page 6
Attachment lli to JPN-92-018 James A. FitzPatrick Nuclear Power Plant Containment isolation Provisions During Statien Blackout EXCLUDED PENETHATIONS Netrunaf Nt/ MARC 97-Penetration Volve 00 Number Sire Gn.)
System Vefve tDs Exclusion SBO Evolustion Cntena 210A 1
RHR Heat Exchanger 10MOV-167A 5
Penetretion escluded besed on e loop seal provvied by a crurumum (cont.)
Vent suppromon pool level.
1 RHR Keep Full 10MOV 95A 5, 3 Merwrnum Flow 3
Core Sprey Pump 14MOV-5A I
Mernmum flow S
Core Sprey Test 14MOV 26A Retu n 1
Core Sprey Holdmg 14 CSP-62 A 3, 5 Pump Merumum Flow 2
2108 16 RHR SPC Test 10MOY-34d Note 3 Penetraton exctuded based on a % seal pmWoo by a erwawmum Retum Une suppresset poc! W.
4 RHR Pump M6rwmum 10MOV-16B Flow Une l
4 RHR Heat Exchanger 10MOV-218 I
Dram 1
RHR Heat Exchanger 10MOV 1678 5
L Vent I
1 RHR Keep Full 10RHR-358 3, 5 Mu mum Flow 3
Core Spray Test 14MOV-58
-i Retum r
A HPCI ?Aremum Flow 23MOV 25 3
- 3. 5 Flow 4
Page 7
4
~
Attcr.hment lll to JPN-92-018 James A. FitzPatrick Nuclear Power Plant Containment Isolation Provisions Ouring Station Blackout EXCLUDED PENETRATIONS Nom'nal NUMARC 87-Penetration Vaive 00 Number Sire (in.)
System Valve ids Exclusion SBC Evaluation Criteria 211A,8 16 RHR to 10MOV-3 S A!B)
Note 2 f: Otrations ncluded based on electrical interlock Suppression Pool preve iting operung of both valves under rermal operattnq Spray 10MOV-39A!BI Note 2 conditions.
212 8
RC1C Turbine 13RCIC-04 3
Penetration esolat+on ensured by check valves 13aCIC-04 Exhaust and 13RCIC-05.
13RCIC45 3
214 20 HP1C Turbine 23HPI-12 3
Penetration isolation cresured by cteck valve 23HPI 12.
Exhaust 23HPl 65 3
217 2
HPCI Turbire 23HPl402 5
Penetration excluded based on 2* nominal size.
~
Exhanst Line 23HPIA03 5
2 yacuum greagg, 2
23MOV 59 5
,~3 218 3/8 Torus Pressure 16-1 AOV-102A 2, 5 Penetration excluded based on 3/8" nommal sire. Th,s Sensing penetrat on is pressure sensing for ILRT Ntrumentation.
16-1 AOV-102B
- 2. 5 220 20 Torus Purge Intet 27AOV-116 2
Penetration excluded based on AOVs failing closed on loss (Air) of a;r.
1.6 Torus Purge inlet 2 7 AOV-132A 5
Penetration excluded based on 1.5* nomenal size.
(Nitrogen) 27 CAD-67 5, 3 27 AOV-1328 5
27 CAD-70
- 5. 3 221 2
RCIC Vacuum 13RCIC47 5, 3 Penetratton escluded based on 2* nominal size.
l Pump Discharge 13RCIC-08 53 222 2
HPiC Turbine 23HP1-13 5, 3 Penetratien ewctoded baseo on 2* nam,nai size.
Drain Trap to Torus 23HPI 56
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w Rv Attachment IV to JPN 92 018 April 1,1992 r
STATION BLACKOUT RULE Containment isolatir n Provisions During a Station Blackout individual Penetration Data Sheets NUTICE: The working sketches in this attachment are provided only as an aid to understanding the applications of the penetration and valve exclusion criteria to specific installations. The sketches and associated tabular data have not been reviewed, approved, or controlledin accordance with procedures for fonnal engineering calculations and drawings.
Note: In the tabular data for each valvo, the last item "SB0 Exclusion," refers to the exclusion criteria 1 throu0h 5 identified in section 7 of NUMARC 87-00 Rev.1 guidelines.
Now York Power Authority James A. FitiPatrick Nuclear Power Plant Docket Number 50 333
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