ML20085C166
| ML20085C166 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/31/1991 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20085C167 | List: |
| References | |
| CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-91-6673, TAC-68546, NUDOCS 9108300153 | |
| Download: ML20085C166 (30) | |
Text
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$AIC 91/6673 a
TECHNICAL EVALUATION REPORT JAMES A. FITZPATRICK NUCLEAR POWER PLANT STATION BLACK 0UT EVALUATION 1
TAC No. 68546 SAIC Science ApplicationsinternationalCorporation An Empwyee owned company Final July 31, 1991 Prepared fort l
L U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract NRC-03-67-029 Task Order No. 38 p --
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TABLE OF CONTENTS Sect 12!1 PJtig-1 1.0 BACKCROUND 3
2.0 REVIEW PROCESS 6
3.0 EVALVATION 3.1 Proposed Station Blackout Duration........
6 3.2 Station Blackout Coping Capability........
10 3.3 Proposed Procedures and Training 18 3.4 Proposed Modifications 19 3.5 Quality Assurance and Technical Specifications 19 20
4.0 CONCLUSION
S 24 S.0 REFERENCES
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TECHNICAL EVALUA110N REPMT JAMES A. FITZPATRICK HUCLEAR POWER Pt>NT SIATION BLACXOUT EVALUATION
1.0 BACKGROUND
On July 21, 1988, the Nuclear Regulatory Comission (HRC) amended its regulations in 10 CFR Part 50 by adding a new section, 50.C?, " Loss of All Alternating Current Power" (1). The objective of this requirement is to assure that all nuclear power plants are capable of withstanding a station blackout (SBO) and maintaining adequate reactor corn cooling and appropriate containment integrity for a required duration.
This requirement is based on information developed under the commission study of Unresolved c fety Issue A-44, " Station a
Blackout" (2-6).
The staff issued Regulatory Guido (RG) 1.155, " Station Blackcut," to provide guidance for meeting the requirements c;f 10 CFR 50.63 (7).
Concurrent with the development of this regulatory guide, the Nuclear Utility Management and Resource Council (NUMARC) developed a document entitled
" Guidelines and Technical Basis for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," NUMARC 87-00 (8).
This docume ?. provides detailed guidelines and procedures on how to assess each plant's
'pabilities to comply with the 5B0 rule. The NRC staff reviewed the guideline and analysis methodology in NUMARC 87-00 and concluded that the NUMARC documentffovides an acceptable guidance for addressing the 10 CFR 50.63 requirements. The application of this method results in selecting a minimum acceptable SB0 duration capability from two to sixteen hours depending on the plant's characteristics and vulnerabilities to the risk from station blackout. The plant's characteristics affecting the required coping capability are:
the redundancy of the onsite emergency AC power sources, the reliability of onsite emergency power sources, the frequency of loss of offsite power (LOOP), and the probable time to restore offsite power.
In order to achieve a consistent systematic response from licensees to the SB0 rule and to expedite the staff review process, NUMARC developed two generic 1
4 response documents.
These documents were reviewed and endorsed (9) by the NRC staff for the purposes of plant specific subWtals. The documents are titled:
1.
" Generic Response to Station Blackout Rule for Plants Using Alternate AC Power," and 2.
" Generic Response to Station Blackout Rule for Plants Using AC Independent Station Blackout Response power."
A plant-specific submittal, using one of the above generic formats, provides only a summary of results of the analysis of the plant's station blackout coping capability. Licensees are expected to ensure that the baseline assumptions used in NUMARC 87-00 are applicable to their plants and to verify the accuracy of the stated results.
Compliance with the SB0 rule requirements is verified by review and evaluation of the licensee's submittal and audit review of the supporting documents as necessary. Follow up NRC inspection:, assure that the lir.ensee has impleinented the necessary changes as requirco to meet the SB0 rule.
In 1989, a joint NRC/SAIC team headed by an NRC staff member performed audit reviews of the meth.odology and documentation that support the licensees' submittals for several plants. These audits revealed several deficiencies which were not apparent from the review of the licensees' submittals using the agreed upon generic response format.
These deficiencies raised a generic question regarding the degree of the licensees' conformance to the requirements of the SB0 rule.
To resolve this question, on January 4,1990, NUMARC issued additional guidance as NUMARC 87 00 Supplemental Questions and Answers (10) addressing the NRC's concerns regarding the deficiencies. NUMARC requested that the licensees send their supplemental responses to the NRC addressing these concerns by Narch 30, 1990.
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4 2.0 REVIEW PROCESS The review of the licensee's submittal is focused on the following areas consistent with the positions of RG 1.155:
d A.
Minimum acceptable SB0 duration (Section 3.1),
B.
SB0 coping capability (Section 3.2),
C.
Procedures and training for SB0 (Section 3.4),
D.
Proposed modifications (Section 3.3), and C.
Quality assurance and technical specifications for 580 equipuent (Section 3.5).
For the determination of the proposed minimum acceptable SB0 duration, the following factors in the licensee's submittal are reviewed:
a) offsite power design characteristics, b) emergency ac power system config *ation, c) determination of the emergency diesel generator (EDG) reliability consistent with NSAC-108 criteria (11), and d) determination of the accepted EDG target reliability.
Once these factors are known, Table 3-8 of NUMARC 87-00 or Table 2 of Regulatory Guide 1.155 provides a matrix for determining the required coping duration.
For the SB0 coping capability, the licensee's submittal is reviewed to assess the availlbti tty, adequacy and capability of the plant systems and components needed to achieve and maintain a safe shutdown condition and recover from an 580 of acceptable durati6n which is determined above. The review pror.ess follows the guidelines given in RG 1.155, Section 3.2. to assure:
a.
availability of sufficient condensate inventory for decay heat
- removal, b.
adequacy of the class lE battery capacity to support safe shutdowr.,
c.
availability of adequate compressed air for air-operated valves necessary fcr safe shutdown, d.
adequacy of the vantilation systems in the vital and/o.' dominant areas that include equipment necessary for safe shutdown of the
- plant, e.
ability to provide appropriate containment integrity, and I
f.
ability of the plant to maintain aequate reactor coolant system inventory to ensure core cooing for the isquired coping duration.
1he licensee's submittal is reviewed to verify that required procedures (i.e., revised existing and new) for coping with SB0 are identified and that appropriate operator training will be providead.
The licensee's submittal for any proposed modifications to emergency AC sources, battery capacity, condensate capacity, compressed air capacity, ventilation system, containment isolation integrity and primary coolant make up capability is reviewed.
Technical Specificaticns and qua*,ity assurance requirements _ set for'h by the license 9 to ensure high reliability of the equipment _ specifically added or assigned to meet the requirements of the SB0 rule, are assessed for their adequacy.
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l This SB0 evaluation is based on a review of the licensee's submittals dated April 17,1989 (12), March 29,19V0 03), the information available in the plant Final Safety Analysis Report (fSAR) (14), a telephone conversation between l
NRC/SAIC and the licensee on January 17, 1991, andthelicensee'sresponse(15) to the questions raised during the telephone conversation; it does not include a cuncurrent site audit review of the supporting documentation.
Such an audit may be warranted as an additi A1 confirmatory action. This determination will be made and the audit may be scheduled and performed by the NRC staff at some later date.
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3.0 EVALUATION 3.1 Proposed Station Blackout Duration Licensee's Submittal The licensee, the New York Power Authority, calculated (12 and 13) a minimum acceptable 500 duration of four hours for the James A. Fitzpatrick (JAF) Nuclear Power Plant.
The licensee stated that some modifications are necessary to attain this proposed coping duration.
The plant f actors used to estimate the proposed SB0 duration are as fc. lows:
1.
Offsite Power Design Characteristics The plant AC power design characteristic group is "P2" based on:
a.
Independence of offsite power group of "11/2 "
b.
Estimated frequency of LOOPS due to severe weather (SW) which places the plant in SW group "3."
c.
Estimated frequency of LOOPS due to extremely severe weather (ESW) which places the plant in ESW Group "1,"
and i
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d.
Expected frequency of grid-related LOOPS of less than once per 20 years, i
l 2.
Emergency AC (EAC) Power Configuration Group l
l The EAC power configuration of the plant is "A."
The JAF plant is equipped with four emergency diesel generators which are normally available to the plant's safe shutdown equipment. One emergency AC S
i we
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..... a power supply is sufficient to operate the minimum required safe shutdown equipment following a loss of offsite power.
3.
Target Emergency Diesel Generator (EDG) Reliability The licensee has selected a target EDG reliability of 0.95 based on having a nuclear unit average EDG reliability of greater than 0.95 for the last 100 demands, consistent with the NUKARC 87 00 criteria.
Review of Licensee's Submittal Factors which affect the estimation of the SB0 coping duration are:
the independence of offsite power system grouping, the estimated frequency of LOOPS caused by grid related failures, the estimated frequency of LOOPS caused by severe weather (SW) and extremely revere weather (ESW) conditicns, the classifiestion of EA(. and the selection of EDG target reliability.
The licensee stated (12) that the NUKARC 87 00 methodology was used for determining the ESW and SW groups.
The licensee's estimate (12) of ESW-caused LOOP frequency is consistent with the information provided in NUKARC 87-00, Table 3-2.
The SW caused LOOP frequency is also consistent with the proper application of NUKARC 87 00, independent of the number of transmission lines rights of way.
The licensee relected an independence of offsite power system grouping of "I 1/2."
A review of the plant FSAR indicates that:
1.
all offsite power sources are connected to the plant through a single switchyard, 2.
the emergency buses are normally powered from the unit main generator through the normal station service transformer (NSST),
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3.
upon loss of power from the main generator, each emergency train is powered from an independent 115 kV offsite power source through a reserve station service transformer (RSST),
4.
any one RSST does not supply loads on both emergency trains for any mode of plant operation, and 5.
upon ' loss of power from one RSST there is no auto, or manual transfer to an alternate power source.
Based on the above, and the guidance provided in Table 5 of RG 1.155, the plant independence of offsite power grouping is "I3."
This determination is based on the RG guidance which states that if the normai source of power to the safe shutdown buses is the unit main generator, there is a need for an automatic transfer and also a manual transfer to the preferred and alternate power sources.
Establishment of the proper Emergency AC (EAC) Configuration Group is based on the number of available EAC sources and the number of EAC sources required to operate safe shutdown equipment following a LOOP.
JAf has four dedicated EAC sources with one required after a LOOP, placing the plant in EAC Group "A" (RG 1.155, Table 3) as the licensee correctly identified.
The final characteristic needed to establish the duration of minimum required coping capability is the target EDG reliability. The licensee stated (12) that the assignment of the EDG target reliability of 0.95 is based on having a unit average EDG reliability of greater than 0.95 for the last 100 demands.
Although this selection is consistent with the criteria given in both the RG 1.155 and NUHARC 87 00, the licensee needs to evaluate the EOG reliability for the last 20 and 50 demands as well.
These statistics are orly available on site for review, therefore, we are unable to verify the assignment of the EDG target reliability at this time. However, based on the information in the NSAC-108, which gives the 7
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, - + -,
w-,e-.~-,,,e e-
EDG reliability data at U.S. nuclear reactors for calendar years 1983 to 1985, the EDGs at JAF experienne an average of 72 demands per year with 100% reliability per diesel per year.
Using tnis data, it appears that the target EDG reliabi'ity selected by the licensee (12) is appropriate.
In response to the requirement for an EDG reliability program the licensee stated, during the telephone conversation on January 17, 1991 and in the
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draft written response to questions (15), that an EDG reliability program ccaistent with the guidance provided in RG 1.155 and NUltARC 87 00 has b n developed, but has nct yet been implemented. The licensee added that a formi program will be imp? mented after the resolution of NRC Generic t
Issue B-56.
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With regard to the expected frequency of srid-related LOOPS at the site, we can not confirm the stated results.
The available information in NUREG/CR-3992 (3), which gives a compendium of information on the loss of l
offsite power at nuclear power plants in the U.S., indicates that JAF did j
not have any sympt-tic grid-related LOOP prior to the calendar year l
1984. (Although it s u dd be noted that this report did indicate that JAF suffered tuo very sho
'Ds due to maintenance errors in 1978 and 1979.
The restoration time was less than five minutes for both events.)
In the absence of any adverse information we agree with the licensee's statement.
Based on an SW group "3," and an ESW group "1," the offsite power design characteristic group of JAF i:
"P2."
With this determination, in conjunction with EAC group "A"
and an EDG reliability of 0.95, the required coping duration for JAF is confirmed to be four hours. Note that l
the determination of the site independence of offsite power grouping does l
not affect the offsite power design characteristic classification.
3.2 Station Blackout Coping Capability l
The piant coping capability with an 560 event for the required duration of four hours is assessed with the following results:
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1.
Condensate Inventory for Decay Heat Removal Licensee's Submittal The licensee stated (12 ) that the JAF FSAR requires a minimum condensate storage tank (CST) level of 200,WO gallons.
The licensee stated that.approximately 54,000 gallons of water are required for decay heat removal for the required copin; dura +ior Of four hours. The licensee stated that no modifications or procedJral changes are necessary to use the CST water source.
Review of Licensse's Submittal For calculating the condensate inventory requirement for JAF during an SB0 event, one has to consider the following:
Using NUMARC 87-00, Section 7.2.1 and the 1.
Decay Heat maximum power level of 2485 MWt, or 102% of nominal power, the plant would require 54,968 gallons of condensate to remove decay heat during four hours.
2.
Reactor toolant Leakaae -- The assumed leak rate of 61 gpm, (a seal leak rate of 18 gpm per recirculation pump, and an additional 25.gpm for the ' maximum technical specification-allowed leakage), results in a total leakage of 14,640 gallons of cc9densate-during four hours.
The sum of the above coolant requiretsnts is $9,608 gallons. This is greater than the licensee's estimate, but much less than the minimum. available CST inventory, Thus, although we do not agree with the licensee's determination of water required during the four hour SB0 coping period, we do agree with the licensee that the site has sufficient condensate-inventory to cope with a 4-hour SB0 event.
Hovever, the licensee needs to-provide assurance that the minimum CST level of 200,00u gallons will be available during normal power 9
operation, in addition, since no depressurization is being considered, the licensee needs to verify that the torus temperature would not exceed its heat capacity temperature limit (llCTL) during an SB0 event.
2.
Class IE Battery Capacity Licensee's Submittal The licensee stated (12) that a battery capacity calculation has been performed in accordance with NUMARC 87-00, Section 7.2.2 to verify that the Class IE batteries have sufficient capacity to meet SB0 loads for four hours.
In response to the questions raised during the telephone conversation on January 17, 1991, the licensee stated (15) that the battery capacity calculations including load stripping were performed in accordance with NUMARC 87-00 and show that either the A or B station battery would be able to power 500 loads for more than six hours.
These calculations assume that the emergency DC lube oil pumps for the main turbine and main feedwater pump turbine, and the non-vital DC lighting are load shed from the A and B station batteries at 30 minutes into an SB0 event.
In addition, at 60 minutes into the SB0 event, the UPS MG set is shed from the A station battery. The shedding of the UPS HG set requires a modification to provide the control room operators with the instrumentation that would only be available in the relay room af ter the load shed.
Review of Licensee's Submittal The OF DC power supply system consists of two 125 V batteries. The plant F:AR, Section 8.7.3, states that the batteries are sized to supply, w! hout recharging, their normal loads for two hours.
The licensee's response to NRC questions (15) states that, with load shedding the batteries can provide power for at least six hours.
Although the licensee stated that the battery sizing calculations 10 l
I assume a low initial electrolyte temperature and end of-life battery condition, we can not confirm the adequacy of the battery sizing calculations.
The licensee was asked during the telephone conversation on January 17,
- 1991, to provide the supporting evaluations for the battery capacity and the effects of shedding the UPS MG set.
In the absence of a response from the licensee, no review of the battery sizing calculation can be performed.
- Thus, the adequacy of the class-lE batteries remain as an open item pending the licensee's response.
3.
Compressed Air Licensee's Submittal The licensee stated (12) that no air operated valves are relied upon to cope with an SB0 for four hours.
The only exception is the reactor building closed loop cooling water (RBCLCW) system isolation valves which fall open upon the loss of instrument air, nitrogen, and DC puwer.
This is the preferred position for in accident condition since it continues to provide emergency service water for drywell equipment cooling in the event of an S80.
Review of Licensee's Submittal in general the instrument air system serves no safety function since it is not required to achieve safe shutdown or to mitigate the consequences of an accident. Air accumulators are provided for each valve where it is required to function for safe shut down of the plant following an accident.
For example, at JAF each of the reactor vessel relief valve; provided for automatic depressurization is equipped with an air / nitrogen accumulator with sufficient capacity to allcw five valve operations. Operation of these valves are required, if reactor depressurization is needed.
It should be noted, however, that the licensee did not consider reactor vessel 11 l
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i depressurization in its calculation of condensate inventory requirement during an SB0 event.
A review of Table 7.3-1 of the JAF FSAR was conducted. This review found that Drywell pressure sensing and Torus pressure sensing penetration valves are normally open and fail closed. The licensee needs to verify that the closure of these valves would not cause the loss of control room pressure indications for these areas.
4.
Effects of Lcss of Ventilation Licensee's Submittal The licensee stated (12 and 15) that the effects of loss ventilation in each areas of concern were determined uilng either a plant-specific thermal transient analysis or the NUMARC 87-00 mathod. The licensee's submittals (12,13 and 15) indicate that the NUMARC 87-00 method was used to calculate area temperature rise for HPCI, RCIC, control room, and steam tunnel. For drywell temperature rise, three plant-specific analyses were used, all of which assumed reactor depressurization.
The licensee reported the following results:
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I AREA TEMPERATURE '
TEMPERATURE (Initial Temp.)
(Ref. 12 )
(Ref,15)
HPCI Room 130*F 130*F (Peak)
(104*F;Ref.15)
RCIC Room 130*F 130*F (bounded by HPCI (104*F;Ref.15) calculation)
Control Room
< 120*F 93*F (75*F ; Ref.15)
Drywell Not 190*F for 65 gpm RPV to drywell f
(Not Provided Provided leak rate after four hours Ref.15)
. Additionally,-the licensee stated that the calculated high SBO.(no actual value provided) main steam tunnel temperature will not cause HPCI or RCIC isolation because SB0 procedure F-A0P-49 directs the J
_ operators to place the affected circuits in the test mode during an SB0 event.
The. licensee stated that reasonable assurance of equipment operability has been assessed using Appendix F to NUMARC 87-00.
Review of Licensee's Submittal It should be noted that the licensee did not submit any room heat-up calculations for review,__only the: calculation results:and some of-the assumed initial conditions for the control room and HPCI analyses (15).
Although discussed during the January-17, 1991 telephone conversav;on, the licensee has not provided the relay room temperature analysis results, an evaluation of the peak SB0 suppression-pool temperature as _ compared to its limit, or further details on the actual calculated SB0 main steam tunnel temperature.
We reviewed the available information in the licensee's SBO-13 6
,.-y
submittals (12 and 15) regarding loss of ventilation evaluations.
This review reveals the following significant comments regarding these analyses:
1.
Based on the review of other BWRs similar to JAF, the peak calculated drywell temperature of 190*F at four hours (after a 4-hour SB0 with an RCS leak rate of 65 gpm), appears to be significantly lower than expected.
The licensee indicated
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during the January 17, 1991 telephone conversation that the drywell design temperature is 309'F (FSAR, Section 5), and that a Browns Ferry's comparison analysis predicts a peak JAF drywell temperature of "... less than 300*F."
The licensee acknowledged that the Browns Ferry's analysis is based on a small RCS leakage. Therefore, this further substantiates that with a 65 gpm leak rate the drywell temperature may approach or exceed the design temperature limit. The licensee needs to review the input parameters used its calculation against the conditions expected during an SB0 event and document the plant spect, ; aspects that would explain why JAF drywell would experience a lower temperature.
2.
The licensee did not perform any suppression pool (torus) heat-up calculation, As with the drywell, the suppression pool has a temperature limit which may be violated during an SB0 event. In addition, the increase in the torus temperature could affect the temperature rise calculation of the adjacent rooms.
The licensee needs to perform an analysis of peak suppression pool temperature during an SB0 with the HPCI being used to cooldown the RCS and heat being deposited in the torus, and evaluate the impact of this temperature rise on the operating equipment in the adjacent areas.
3.
The licensee did not provide the technical details or the calculation results for the main steam tunnel SB0 heat-up
- analysis, in addition, the licensee did not address the 14
d f
operability of HPCI and RCIC turbine steam supply valves in the elevated temperature environment of the main steam tunnel, should that be needed to provide adequate containment integrity.
4.
The licensee did not address room heat-up in the relay room.
This room needs to be evaluated.
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5.
The assumed initial control room temperature of 75*F is non-conservative unless the licensee has an appropriate administrative controls to ensure that under no circumstances this temperature would be exceeded.
on the other hand, the licensee calculated a final room temperature of 93*F using the 75 F initial temperature.
If other assumptions used in the heat-up calculation are adequate, then the initial temperature can be as high as 102*F before it would cause the room to be a dominant area of concern.
Thus, pending verification of this analysis, the control room is not a dominant area of concern.
The licensee, however, needs to have a procedure which calls for opening the control room cabinet doors within 30 minutes of an SB0 event consistent with the guidance.
Based on the above coments, the licensee's heat-up calculations are considered to be incompTete requiring revision or re-analysis as necessary.
Until such time, this issue remains open pending NRC's review of licensee's response.
5.
Containment Isolation Licensee's submittal The licensee stated (12) that containment irolation valves (CIVs) that must be captble of being closed or that must be operated (cycled) under SB0 conditions can be positioned (with indication) independent of the preferred and blacked-out unit's Class IE power 15
l supplies.
No actions and/or procedure changes were required to provide appropriate containment integrity during an SB0 event.
The licensee added (13 and 15) that in lieu of the five exclusion criteria given in NUMARC 87-00 and RG 1.155, a different approach was used.
The licensee stated that each line penetrating containment is isolated by a minimum of ona:
1.
fail-closed valve, l
2.
locked closed valve, I
3.
check valve, 4.
valve interlocked with another valve in the same penetration, or 5.
DC powered motor-operated valve.
The licensee stated that this method of evaluating each containment penetration to the aforementioned criteria fulfills the purpose, and is consistent with NUMARC 87-00.
Review of Licensee's Submittal Our review of the containment penetrations and the associated CIVs was limited by insufficient information in Section 7.3 and Table 7.3-1 of the JAF FSAR.
The methodology described by the licensee was compared to that in NUMARC 87-00 and RG 1.155. although similar in intent, the licensea's approach inchoes one criterion that is not in conformance with the guidelines established by RG 1.155 and NUMARC 87-00.
This criterion is a valve which is interlocked with l
another valve in the same penetration. The licensee didn't identify which CIVs/ penetrations are excluded using this criterion.
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'1 Therefore, the licensee needs to add the valves that are excluded by the additional criteria in an appropriate procedure and identify actions which are needed to confirm these valves are closed, if needed.
The valve closure needs to be confirmed by position indication, (local, remote, mechanical, process information, etc.).
6.
Reactor Coolant Inventory
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Licens'ee's Submittal The licensee stated (12) that the ability to maintain adequate reactor coolant inventory was assessed in a plant-specific analysis.
The analysis shows that expected rates of reactor coolant inventory loss under SB0 conditions do no result in more than a momentary core uncovery for an SB0 of four hours.
Therefore, the licensee concluded that makeup systems in addition to those currently available under SB0 conditions are not required.
Review of Licensee's Submittal Reactor coolant make-up is necessary to remove decay heat, cool down the primary system, and to replenish the RCS inventory losses due to the recirculation pump seal leakage (18 gpm per pump per NUMARC 87 00 guideline) and the maximum allowable technical specification leakage (estimated to be 25 gpm). With no cooldown, the RCS loses 61 gpm, and on the average boils off 230 gpm ror a total inventory loss of 291 gpm.
Both RCIC and HpCI have sufficient capacity to provide the needed RCS make-up to maintain the core cooled and covered.
NOTE:
"The IB com recirculation pumo seal leak rate was agreed to between NUMARC and the staff pending resolution of generic Issue (GI) 23.
If the final resolution of GI-23 defines 17 l
I higher seal leak rates than assumed for the RCS inventory evaluation, the licensee needs to be aware of the potentia!
impact of this resolution on its analyses and actions addressing conformance to the SB0 rule."
3.3 Proposed Procedures and Training Licensee's Submit *,al The licensee stated that plant procedures have been reviewed and modified to meet the guidelines in NUMARC 87-00, Section 4 in the following areas:
1.
Station Blackout, and 2.
Severe Weather.
The licensee added that the following plant procedures have been reviewed and changes to meet NUMARC 87-00 will be implemented:
1.
Station Blackout Response, and 2.
Procedure changes required to reflect the needed modifications.
Review of Licensee's Submittal We neither received nor reviewed the affected procedures or training.
These procedures are plant specific actions concerning the required activities to cope with a SB0.
Our cursory review indicates that the licensee did not consider the plant AC power restcration procedure requiring any review or modificatica(s). It is licensee's responsibility to review, revise and implement the affected procedures, as needed, to mitigate an SB0 event and to assure that these procedures are complete and correct in their contents and that the associated training needs are carried out accordingly.
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3.4 Proposed Modifications Licensee's Submittal The licensee stated (12 and 15) that some modifications would be required to cope with an SB0 with a curation of four hours.
These modifications are delineated below.
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1.
Modify RCIC enclosure vent fan power source to a qualified DC.
2.
Supply the control room instrument panel 27 HAP with an emergency electrical (DC) power source.
Review of Licensee's Submittal The licensee's proposed modifications are in accordance with the applicable guidance of RG 1.155 and NUMARC 87-00.
If properly implemented, the modifications to the RCIC system will reduce the RCIC room temperature during an
- SBO, and the modification to the instrumentation panel power supply will improve the operator's information during an SB0. Our review, however, has identified several concerns which the licensee needs to respond and which may require additional modifications for their resolutions.
3.5 Quality Assurance And Technical Specifications In response to the questions raise during the telephone conversation on January 17, 1991 the licensee stated (15) that, with the modification to the RCIC enclosure ventilation fan power supply, ali equipment which is relied upon to cope with a postulated SB0 event will be Quality Assurance (QA) Category 1.
The licensee did not address any specific technical specification requirements associated with 580 coping.
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0 4.0 CONCt.USIONS Based on our review of the licenseef s submittals, a telephone conversation between NRC/SAIC and the licensee, and the information available in the UFSAR for the James A Fitzpatrick Nuclear Power Plant, we find the submittal conforms with the requirements of the SB0 rule and the guidance of RG 1.155 with the following exceptions:
1.
Independence of Offsite Power System The licensee classified the site independence of offsite power system as "11/2." Our review of the plant offsite power system and its connectability and capability indicate that it should be classified as "13."
This determination is based on the RG 1.155 guidance which states that if the normal source of power to the safe shutdown buses is the unit main generator, there is a need for an automatic transfer and also a manual transfer to the preferred and alternate power sources. The plant has only one automatic transfer to the preferred power source.
However, this determination does not affect the offsite power characteristic or the minimum required duration.
2.
Condensate Inventory Although we agree with the licensee that the site needs $uch less than the that available in the CST's for decay heat removal during an SB0 event, the licensee needs to ensure that the minimum CST volume of 200,000 gallons will be available during plant operation.
3.
Class 1E Battery capacity Review of the plant FSAR indicates that the Class lE batteries are sized to supply their normal loads for two hours without recharge.
The licensee stated that these batteries can support the SB0 loads for more than six hours with load shed.
The licensee did not 20 1
provide the supporting evaluations for the battery capacity and the proposed modification resulting from the shedding of the UPS MG set.
Thus we didn't review the battery sizing calculations.
4.
%npreesed air Our review of the plant FSAR indicates that Drywell pressure sensing and Torus pressure sensing penetration valves are normally open and fail closed. The licensee needs to verify that the closure of these valves would not cause the loss of pressure indications for these areas in the control room.
5 Effects of Loss of Ventilation a.
DrYwell Heat-u2 The licensee calculated a peak drywell temperature of 190*F for a four hour SB0 and a 65 gpm RCS leak rate.
In comparison to other similar BWRs, this temperature appears to be significantly lower than expected.
The licensee needs to review the input parameters used its calculation against the conditions expected during an SB0 event and document the plant specific aspects that would explain why JAF drywell would experience a lower temperature.
b.
Torus (Suporession Pool)
The licensee did not perform or report results of any suppression pool heat-up analysis during an SB0 event.
The torus is used as a heat sink during decay heat removal and cooldown, if attempted, which occur during an SB0 event. Torus has a pool temperature limit which must not be violated to ensure that the pool can maintain its function. In addition, higher temperature in the torus would affect the adjacent room temperatures. Thus, the licensec needs to perform
- t. suitably conservative calculation of suppression pool heat-up 21
/j C
during an SB0 event, and evaluate the potential impact of this t 7perature on the operating equipment in the adjacent areas.
c.
Main Steam Tunnel Although the licensee qualitatively discussed a main steam tunnel heat up calculation and addressed a procedure revision to avoid HPCI and/or RCIC pump trip from high temperatures in this room, specific information on the calculated peak temperature for this room has not been provided.
Al so, the licensee needs to ensure that all equipment (i.e. the HPCI/RCIC turbine steam supply valves) in this room which is required for SB0 coping is operable in the calculated temperature environment, d.
Relay Room The 1Icensee needs to perform and report the results of a room heat-up calculation for the relay room.
This room contains equipment required for SB0 coping.
e.
Control Room The assumed initial control room temperature of 75'F is non-conservative unless the licensee has appropriate technical specifications and/or administrative controls that would ensure that this temperature would not be exceeded under any circumstances. Our review indicates that, with the final calculated temperature of 93*F using the low initial temperature, the room has sufficient temperature margin before it becomes a DAC provided that all other assumptions used in the calculations are adequate.
Pending such verification, we consider the initial room temperature assumption to be inconsequential. However, the licensee needs to have a procedure which calls for opening the control room cabinet doors within 30 minutes consistent with the guidance.
22
~
y
- 6. _
Containment Isolation The licensee approach for excluding CIVs uses additional criteria to those given in'RG 1.155'and'NUMARC 87-00,-(see_ item 5 in Section 3.2).
Therefore, the licensee needs to add the. valves that are excluded by the additional criteria in an appropriate procedure and _
identify ac.tions which are needed to confirm these valves are closed, if needed.
The valve closure needs to be confirmed by position indication, (local,-
- remote, mechanical, process information,.etc.).-
/
7.
Proposed Modifications Our review has identified several concerns which the licensee needs to respond and which may require additional modifications for their resolutions.-
F l
l 23
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-
2tt '
r-
5.0 REFERENCES
-1.
The Office of Federal Register,."Ccde r; f?4eral Regulations Title 10 Part 50.63," 10 CFR 50.63, January 1, 1969.
2.
U.S.
Nuclear Regulatory Comission, Evaluation of Station Blackout Technical Findings Related to Accidents at Nuclear-Power Plants Unresolved Safety issue A 44," NUREG-1032, Baranowsky, P. W., June 1988.
3.
U.S. Nuclear' Regulatory Comission, " Collection and Evaluation of Complete and Partial-Losses of Offsite Power at Nuclear Power Plants," NUREG/CR-3992, February 1985.
4.
U.S. Nuclear Regulatory Comission, " Reliability of _ Emergency AC Power
-System at Nuclear Power DLnts," NUREG/CR-2989, July _-1983.
5.
U.S. Nuclear Regulatory Comission, " Emergency Olesel Generator Operating
- Experience,'1981-1983," NUREG/CR-4347, December 1985.
6.
U.S. Nuclear Regulatory Comission, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," NUREG/CR-3226, May 1983.
7.
-U.S. Nuclear Regulatory Comission Office of Nuclear Regulatory.Research,
" Regulatory Guide 1.155 Station Blackout," August 1988.
8._
. Nuclear Management and Resources Council, Inc., " Guidelines and Technical Basesifor NUMARC' Initiatives Addressing Statir.n Blackout-at Light Water Reactors," NUMARC 87 00, November 1987.
9.
Thadani,- A. C., Letter to W.
.H.
Rasin of - NUMARC, " Approval of NUMARC Documents on Station Blackout _(TAC-40577),"- dated October 7,1988.
10.
- Thadani, A.' C.,-letter to A. Marion of NUMARC, " Publicly Noticed Meeting DecemberL27, 1989," dated January 3, 1990.
24-v-
.r-
--r.
4 er-
,w.,
=
444
- 11. -
Nuclear. Safety Analysis Center, "The Reliability of Emergency Diesel Generators at U.S. Nuclear Power Plants," NSAC-108, Wyckoff, M., September 1986.
12.
Brons, J.
C.,
letter to U.S.
Nuclear Regulatory Commission Document 7
Control Desk, " James A. Fitzpatrick Nuclear Power Plant, Docket No: 50-333 Response to 10 CFR 50.63, Loss of All Alternating Current Power -
Station Blackout", JPN-89-018, April 17, 1989.
13.
Brons, J.
C.,
letter to U.S. Nuclear Regulatory Comission Document Control Desk, " James A. Fitzpatrick Nuclear Power Plant, Docket No: 50-333, Station Blackout Supplementary Information", JPN-90-026, March 29, 1990.
14.
James A. - Fitzpatrick Nuclear Power Plant, Updated Final. Safety Analysis Report.
15.
James A. Fitzpatrick Nuclear Power Plant, " Responses to NRC Request for Additional Information Regarding Station Blackout", Attachment I to JPN-91-DRAFT, January 17, 1991.
25
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,.,w.-
.O A
Mr. Ralph E. Beedle November 13, 1991 adequately addressed in your September 13, 1991, letter.
If this is the case, simply reference that letter in response to the specific recommendation.
All analyses, confirmations, and ether documentation supporting your SB0 submittals should be maintained and available for further NRC staff inspection and assessment. The NRC staff is currently cons'dering Technical Specifications (TS) for SB0 equipment in context of the TS Improvement Program.
In the interim, plant procedures to reflect the appropriate testing and surveillance requirements should be in place to ensure the operibility of the necessary SB0 equipment.
You will be notified if a determination is made that TS are required for SB0 equipment.
This requirement for confirmation and information affects one respondent; therefore, is not subject to Office of Management and Budget review under P.L.96-511.
Sincere ly,
Original Signed By:
Brian C. McCabe, Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Enclosure:
Safety Evaluation cc w/ enclosure:
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