ML20091J115

Final Technical Evaluation Rept,Susquehanna Steam Electric Station Units 1 & 2,Station Blackout Evaluation

ML20091J115
Person / Time
Site:	Susquehanna
Issue date:	10/01/1991
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML17157B019	List: ML17157B018 ML17157B020 ML20091J115
References
CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-91-6671, TAC-M68613, TAC-M68614, NUDOCS 9201060035
Download: ML20091J115 (36)

Text

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TEC11NICAL EVALUATIC,N REPORT SUSQUEllANNA STEAM ELECTPlc STATION UNITS 1 & 2 STATION BLACKOLTr EVALUATION TAC Nos. 68613,68614 5AIC Final October 1,1791 Prepared fon U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract NRC 03 87 029 Task Order No. 38

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i TABLE OF CONTENTS 1

Section Eagg l

1.0 B A C KG R O UN D................................................................ 1 2.0 R E V 1 EW PR OC E S S............................................................. 3 t

3.0 E V A L U A T I ON......................................................................... 5 3.1 Proposed Station Blaekout Duration........................ 5 3.2 Station Blackout Coping Capability........................... 12 3.3 Propostd Procedures and Training.........................

27 3.4 Proposed M odific a tions...............................................- 27 3.5 Quality Assurance and Technical Specifications.... 28 4.0 C ON C LU SION S........................................................................ 2 9 5.0 RE FE RE NC E S....................................................................... 3 2 t

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TECHNICAL EVALUATION REPORT SUSQUEtIANNA STEAM ELECTRIC STATION UNITS 1 & 2 STATION BLACKOUT EVALUATION L

1.0 BACKGROUND

On July 21,1988, the Nuclear Regulatory Commist,lon (NRC) amended its regula"ons in 10 CFR Part 50 by adding a new section,50.63," Loss of Al' Alternating Current Power"(1) The objective of this requirement is to assure that all nuclear power plants are capable of withstanding a station blackout GBO) and maintaining l

adequate reactor core cooling and appropriate containment integrity for a required c'uration. This requirement is based on information developed under the commission study of Unresolved Safety Issue A 44," Station Blackout" (2 6).

1 The staff issued Regulatory Guide (RG) 1.155, " Station Blackout," to provide guidance for_ meeting the ret,uirements of 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 document provides detailed guidelines and procedures on how to assess each plant's capabilities to comply with the SBO rule. The NRC'siaff reviewed the guidelines and analysis methodology in NUMARC 87 00 and concluded that the NUMARC ciocument provides an acceptable guidance for addressing the 10 CFR 50.63 requirements.- The application.

of this method results in selecting a minimum acceptable SBO 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 SBO g

l rule and to expedite the staff review process, NUMARC developed two generic p

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n response documents. These documents were reviewed and endorsed (10) by the NRC staff for the purposes of plant specific submittals. 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 SBO 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 inspections assure that the licensee has implemented the necessary changes as required to meet the SBO rule.

In 1989, a joint NRC/SAIC team headed by an NRC staff member performed audit reviews of the methodology 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 licensees' conformance to the requirements of the SBO rule. To resolve this question, on January 4,1990, NUMARC issued additional guidance as NUMARC 87 00 Supplemental Questions / Answers (11) addressing the NRCs concerns regarding the deficiencies. NUMARC requested that the licensees send their supplemental responses to the NRC addressing these concerns by March 30, 1990.

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2.0 REVIEW PROCESS The review of the licensee's submittalis focused on the following areas consistent with the positions of RG 1.155:

A.

Minimum acceptabie 500 duration (Section 3.1),

B.

SBO coping capability (Section 3.2),

C.

Procedures and training for SBO (Section 3.3),

D.

Proposed modifications (Section 3.4), ud E.

Quality assurance and technical specifications for SBO equipment (Section 3.5).

For the determination of the proposed minimum acceptable SBO duration, the following factors in the licensee's submittal are reviewed: a) offsite power design characteristics, b) emergency AC power system configuration, c) determination of the emergency diesel generator (EDG) reliability consistent with NSAC 108 criteria (9), 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 RG 1.155 provides a matrix for determining the required coping duration.

For the SBO coping capability, the licensee's submittal is reviewed to assess the availability, adequacy and capabi' ty of the plant systems and components J

needed to achieve and maintain a safe shutdown condition and recover from an SBO of acceptable duration which is determined above. The review process 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 1E battery capacity to support safe shutdown, c.

availability of adequate compressed -ir for air operated valves necessary for safe shutdown, 3

d.

adequacy of the ventilation systems in the vital and/or dominant areas that include equipment necessary for safe shutdown of the plant, ability to provide appropriate containment integrity, and e.

a f.

atrility of the plant to maintain adequate reactor coolant system inventory to ensure core cooling for the required coping duration.

The licensee's submittal is reviewed to verify that required procedures (i e.,

revised existing and new) for coping with SBO are identified and that appropriate operator training will be provided.

The 1 ensee's submittal for any proposed modifications to emergency AC sources, battery capacity, con 'ensate capacity, compressed air cc.pacity, appropriate containment integrity and primary coolant make up capability is reviewed.

-Technical specifications and qual;ty assurance set forth by the licensee to ensure high reliability of the equipment, specifically added or assigned to meet the requirements of the SEO rule, am assessed for their adequacy.

This preliminary SBO evaluation is based upon the review of the licensee's submittals dated April 17,1989 (13), April 17,1990 (14), and February 27,1991 (15), the licensee's written response (17) to questions discussed at the June 14,1991 telephone conference, and the information available in the plant Final Safety Analysis Report (FSAR) (12);it does not include a concurrent site audit review of the supporting documentation. Such an audit may be warranted as an additional confirmatory acticn. This determination would be made and the audit would be scheduled and performed by the NRC staff at some later date.

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e 3.0 EVALUATION 3.1 Proposed Station Blackout Duration Licensee's Submittal The licensee, Pennsylvania Power and Light Company, calculated (13) a minimum acceptable station blackout duration of four hours for the Susquehanna Units 1 and 2. The licensee stated that no modifications were required to attain this coping duration.

The plant factors used to estimate the proposed SBO duration are:

1.

Offsite Power Design Characteristics The plant AC power design characteristic group is "P1" based on:

Independence of the plant offsite power system characteristics of a.

"11/ 2,"

b.

Expected frequency of grid related LOOPS of less than one per 20

-years, c.

Estirnated frequency of LOOPS due to extremely severe weather (ESW) which places the plant in ESW Group "2," and d.

Estimated frequency of LOOPS due to severe weather (SW) which places the plant in SW Group "2," and 2.

Emergency AC (EAC) Power Configuration Uroup i

in its original submittal, the licensee stated (13) that the EAC power configuration group is "D," bued on:

There are four emergency power supplies not credited as alternate AC sources.

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Three emergency AC power supply are necessary to operate safe shutdown equipment following a loss of offsite power.

However, in a later submittal, the licensee revised its position and stated (15) that only two emergency power supplies are necessary to operate safe shutdown equipment following a loss of offsite power, and thus, the EAC power configuration group is "It" 3.

Target Emergency Diesel Generator (EDG) Reliability In its original submittal, the licensee selected (13) a target EDG reliability of 0.975. The selection of this target reliability was based on having a nuclear unit average EDG reliability for the last 100 demands greater than 0.95, consistent with NUMARC 87-00, Section 3.2.4.

In a later submittal, however, the licensee revised its position and reduced (15) the minimum required EDG reliability target value from 0.975 to 0.95.

Review of Licensee's Submittal Factors which affect the estimation of the SBO coping duration are: the independence of the offsite power system grouping, the estimated frequency of LOOPS due to ESW and SW conditions, the expected frecuency of grid-related LOOPS, the classification of EAC, and the selection of EDG target reliability, Using Table 3-2 of NUMARC 87-00, the expected frequency of LOOPS due to ESW conditions place the Susquehanna site in ESW Group "3." In response to questions raised in the June 14,1991 telephone conference, the license stated (17) that it used Thom's method (20) and calculated a 1,000 year return period " fastest mile" wind speed of approximately 110 mph at a height of 30 feet above the ground. Further, the licensee stated that for a 125 mph wind speed the return period would be 1/1,500 years or 6.67E 04.

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The calculation performed by the licensee was normalized to a height 30 feet above the ground, while the NRC data listed in NUMARC 87 00 Table 3 2 is based on a normalized height of 30 meters which represents the average l

distribution line height. A 104-mph wind speed normalized to 10 meters is equivalent to a 125 mph windspeed normalized to 30 meters with each giving a return period of 1/1,000 years or 1E 03. Since the normalized wind speed reported by the licensee at a height of 30 feet exceeds 104 mph, we conclude, based on the licensee's data, that the return period for winds greater than 125-mph normalized to a height of 30 meters is greater than 1E 03 for the Susquehanna site, and thus, the Susquehanna site is in ESW Group "3."

Using data from Table 3 3 of NUMARC 87-00, the expected frequency of LOOPS due to SW conditions place the Susquehanna site in SW Group "2,"

which is in agreement with what was stated in the licensee's submittal (13).

This calculation was performed assuming there were multiple rights of way among the incoming transmission lines, consistent with Figure 8.21 of the plant FSAR (12).

The licensee stated that the independence of offsite power system grouping is "11/2." A review of the Susquehanna FSAR (12) shows that:

1.

There are two electrically connected switchyards for the site; 2.

During normal plant operation, each unit's emergency busses are powered from two 230 kV offsite power sources though the start up transformers (SUTs), as depicted in Figure 1; 3.

Each SUT feeds two engineering safeguards transformers (ESTs);

4.

Each EST is connectable to four emergency b*usses in both units, two of which are normally connected;

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Upon a loss of power from one of the SUTs, the effected emergency busses will automatically be powered by the second SUT from automatic transfers at the 4.16 kV emergency bus level (see Figure 1).

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l j a Based on the above, the plant irdependence of offsite power system group is "12.* This determination is based on the guidance of Table 5 of RG 1.155. With regard to EAC classification, the current position of the licensee is that Susquehanna is in group "B," based on the assertion that two out of four available EDGs are required tr: achieve and maintain a safe shutdown for both l units. We have reviewed the applicability of the above c; aim by considering both EDG capability and cornectability as detailed below. CaPaM.lty.i Bared on the EDC load list provided in FSAR Table 8.3 2, the total load requirement to bring both units to a safe shutdown following a LOOP is estimated at approxJmately 7913 kW. For two EDGs each needs to carry 3957 kW. According to the plant FSAR, each of the diesels are rated at 4000 kW for continuaus operation and 4700 kW for 2000 hours of operation, thus, it appears that the licensee's assertion that two EDGs are required to maintain a safe shutdown following a LOOP is valid. Conngetability: The licensee stated (17) that manual actions will be required if only two of four FDG are used to cool down both reactors simultaneously following a ' OOP. There are six unique combinations of two EDGs that can be used to bring the units to a safe shutdown condition. For four of the combinations (A&B, A&D, B&C, C&D) one RHR pump vrill be ava!!able in each unit for cooldown, howcVer, remote manual operation of one of the RHR SPC. valves will be required in each unit. For the remaining two combinations of EDGs (A&C, B&D) only one RHR pump is effectively available to cooldown both l units, since only one division of ESW is available in each case. However, the RHR pump powered by the available EDGs in each case requires both divisions of ESW pumps for cooling. The licensee stated that a single RHR pump is sufficient to bring both units to cold shutdovm through " staggered l 9 (

operation" of the RHR pumps in which cooling is manually shifted between s units. Our review of FSAR Table S.31 reveals that the control building structure cooling, ventila; ion and heating for both units are powered from the C and D EDGs. Therefore,if A and B are the available EDGs, there will be no HVAC available at the control building structure. Since EAC classification is based on loads which support the safe shutdown operation of the plant for an extended period, as stated in NUMARC 87-00 Supplemental Questions and Answers the combination of the A and B EDCs does not rneet the connectability criterion.1urther, to provide a full division of AC and DC support, the plant needs to have EDCs A and C or B and D available. In all other combinations, AC power is available for the extended period but a fall division of DC power may not be available for the extended period. For the combinations of A and C and B and D, the dependency of the RHR and ESW pumps requires staggering the operation of the RHR pumps between units to achieve a safe shutdown consistent with the guidance. However, this action is not consistent wi'.h the staff position which states (18) that any actions that would add to the burden of operators that are already in a high stress environment, such as load switching, are considered to be a degradation of the normal safe shutdown capability for a LOOP. Thus, we conclude that three EDGs are required to power the LOOP shutdown loads for both units. The three-out-of four EDG requirement results in a site EAC power configuration value of "D." The licensee initially selected (13) an EDG target reliability of 0.975 based upon having a nuclear urut average EDG reliability greater than 0.95 for the last 100 demands. However, in a later submittal, the licensee revised its position and reduced (15) the minimum required EDG reliability target value from 0.975 to 0.95. Although the information used by the licensee is an acceptable criterio-for choosing an EDG target reliability, the guidance of RG 1.155 requires thr.t the EDG statistics for the last 20 and 50 demands also be calculated. Without I I this information, it is difficult to judge how well the EDGs have performed in the plt and if there should be any concern. We are unable to verify the demonstrated start and load-run reliability of the plant EDGs. This t l information is only available onsite as part of the submittal's supporting i 10 .-c-- c---. m-

documents. The information in NSAC-108, which gives EDG reliabilit) lata at US nuclear power plants for the calender years 1933 to 1985, indicates that the EDGs at Susquehanna experience an average of 90 valid start demands per diesel calender year and have reliability levels of higher thar 0.99. Using this data,it appears that the licensee's selection of the EDG target reliability (0.95) is appropriate. Nevertheless, the licensee needs to have an analysis showing the EDG reliability statistics for the last 20,50 and 100 dernands in its SBO submittal supporting documents. In response to questions regarding EDG reliability program rained at the June 14,1991 telephone conference, the licensee stated (17) tnat PP&L is in the process of implementing a reliability program consistent with the elements provided in RG 1.155, Position 1.2, which includes: individual EDG reliability target levels, surveillance testing and reliability monitoring, an EDG maintenance program, an information and data collection system, and a management oversight progren. With regard to the expected frequency of grid-related LOOPS at the site, we can not confirm the stated results. The available information in NUREG/CR-3992 (3), which gives a compendit:ra of information on the iass of offsite power at nuclear power plant in the U.S., indicates that Susquehanna had not experienced any partial or significant losses of offsite power prior to the calender year 1984. In the absence of any contradictory information, we agree with the licensee's statement. Based on an ESW group "3," SW group "2," and an EAC classification "D," the offsite power design characteristic of the Susquehanna site is "P2" requiring an EDG target reliability of 0.975, with a minimum coping duration of eight hours. This is different from the "P1" offsite power design characteristic and four hour minimum coping duration reported by the licensee. In the following sectii as, we have reviewed the plant coping i capability for a duration of eight nours. 11 L ~. ~,_

3.2 Station Blackout Coping Capability The plant coping capability with an SBO event for a required duration of eight hours is assessed with the following results: 1. Condensate Inventory for Decay Heat Removal Licensee's Submittal The licensee stated (13) that 117,626 gallons of water are required for the decay heat removal during the four hour coping period. The minimum permissible Condensate Storage Tank (CST) level per Technical Specifications provides 135,000 gallc,ns of water, which exceeds the required quantity for coping with a 4 hour SBO event. In response to questions raised at the June 14,1991 telephone conference, the licensee provided the contributing elements to its calculation of condensate inventory as follows: PP&L Method: Decay Heat = 55,000 gallons RCP sealleaks = 19,700 gallons Blowdown = 18,100 gallona 93,300 gallons required for 4 hours NUMARC 87-00 Method: Decay Heat = 72,841 gallons RCP sealleaks = 22,500 gallons Blowdown = 22,285 callons 117,626 gallons required for 4 hours In its coping assessment the licensee stated (19) that it has performed its own in house analysis which assumed a constant reactor vessel leakage of 100 gpm starting 15 minutes into the SBO event and reactor 12

depressurization to 200 psia using one SRV. Based on these conditions, the licensee concluded that the minimum CST level will provide enough water to maintain the reactor vessel water level constant f< six hours. Under the same conditions, however, the licensee stated that ths manual connection of the RWST to the CST would extend the core cooling capability of the CST to more than 20 hours. Review of 1.icensee's Submittal Using the expression provided in NUMARC 87 00 (assuming no cooldown), we estimated that 122,256 gallons of water per unit would be required to remove decay heat during an eight hour SBO event. This estimate is based on the maximum licensed core thermal rating of 3439 MWt per unit listed in the Susquehanna FSAR (12). Assuming a reactor coolant system leakage of 61 gpm (18 gpm per reactor recirculation pump and 25 gpm technical specification leakage) would result in an additional 29,280 gallons in losses. In addition, the licensee calculated that rm another 22,285 gallons would be required for blowdown. This results in a total condensate requirement of approximately 174,000 gallons which exceeds the minimum permissible Condensate Porage Tank level of 135,000 gallons of water. According to FSAR Table 6.2-1 and 6.2-4, there appears to be a t minimum of 960,000 gallons of condensate available in the Suppression pool at a temperature of 90'F at the beginning of the SBO event. However, in its coping analysis, the licensee states (19) that its SBO coping strategy recommends that the HPCI suction transfer from the CST to the suppression pool because the high suppression pool level will be bypassed within the first 75' minutes of the SBO event.- If the suct!3n transfer does occur, the HPCI pump may fall on high lube oil temperature or high pump seal water temperature before the end of the eight hour SBO event. In addition,in its coping assessment the licensee stated (19) that there is a minimmn of 270,000 gallons of water in the RWST which can be manually connected to refill the CST during an SBO. 13

. _. ~. - .. _ _ _ _ -. -. _. _ -. _ _.. _.__ _ _ _ _ _._ _._.~.._ Based upon information provided by the licensee and the results of our, review, we condude that the CST has insufficient capacity to provide adequate core cooling for the entire eight hour coping duration. Thus, the licensee needs to provide for adequate core cooling for the entire eight hour SBO duration either by proceduralizing the manual connection of the CST to the RWST, or by utilizing the cool water available in the Suppression pool at the onset of SBO, to supplement or replace the CST and ensure the availability of the high pressure injection systems. 2. Class 1E Battery Capacity Licensee's Submittal The licensee stated 03) that a battery capacity calculation has been l performed which verifies that the station Class 1E batteries 025V and 250V) have sufficient capacity to meet 5BO loads for four hours assuming loads not needed to cope with a SLO are tripped. The licensee further stated that these loads are identified in plant procedures. In its SBO coping assessment, the licensee stated 09) that the current 125V DC battery cells will be upgraded from nine posith plates to ten positive plates following each unit's refueling outage later this year. Based on these battery modifications, the licensee made the following design assumptions to develop load profiles and calculate battery endurance: A temperature correction factor of 1.116ased on a minimum expected electrolyte temperature of 60'F was used. An aging factor of 1.25 was used. To allow for additional loads, the actual load was increased by two amperes and rounded up to the next integer. 14 _ -...,. _ ~ _, _ - - - -..

The following non class 1E loads need to be manually shed from the 250 VDC batteries (ID650 and 1D660) to extend battery capacity to four hours: A. Reactor Feedwater Pumps Emergency Lube Oil Pumps B. Reactor Recirculation MG Set Emergency Lube Oil Pumps C. Turbine Cenerator Emergency Lube Oil Pump D. Turbine Generator Emergency Seal Oil Pump E. Computer UPS 656 The battery endurance calculations performed by the licensee are summarized as follows: Channel / Final Terminal Battery Division Endurance (hrs) Voltage ...................................n................. ID610 1A/I 6.8 108.3 ID620 IB/II 6.4 108.7 1D630 1C/I 13.2 108.0 1D640 1D/D 12.2 108.2 1D650 I 7.8

• 1D660 11 4.3
• 2D610 2A/1 6.3 108.8 2D620 2B/II 5.9 109.0 2D630 2C/I 11.3 108.1 2D640 2D/II 10.8 107.9 2D650 1

9.3 2D650 11 6.4 ' Includes Load Shedding .....................==............................. The licensee ussi the results of the battery endurance calculations to demonstrate thal the batteries have sufficient capacity to supply SBO loads for a coping duration of four hours. In addition, the licensee stated (19) that a portable AC generator has been installed which is i 15

designed to provide AC power to the "A" and "B" channel battery chargers in the event of an extended SBO, The licensee estirnates that the generator can be connected within one to two hours to the plant AC distribution system. Once tied to the battery chr.rgers, the diesel is expected to function for at least forty hours. In addidon, the licensee stated (16) that a plant modification to install a non class 1E batte.y to carry all non class IE loads, thus permanently i shedding these loads from batteries 1D650 and ID660 ensuring at least four hour capacity,is under evaluation. In response to questions raised at the June 14,1991 telephone conference, the licensee provided (17) a copy of its battery capacity calculations for 125 VDC batteries ID610 and 1D630 for review. Review of Licensee's Submittal The batteries should be able to provide the normal plant monitoring and control for the entire SBO duration of eight hours. According to the Susquehanna FSAR, the design basis for battery sizing is four hours. Based on a review of the licensee's battery capacity calculations for SBO loads, we conclude the following:

• The licensee's assumed temperature correction factor of 1.11 (based on an electrolyte temperature of 60*F) and agmg factor of 1.25 are conservative and consistent with NRC guidance.
• The licensee did not use a design margin in its calculation. This is not consistent with the recommendation of IEEE Std 485, which states a 10% to 15% design margin needs to be considered.
• The licensee's assumed SBO load profile is consistent with the information contained in the FSAR.

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• The !!censee did not assume any diesel generator field flash attempts as part of its first minute random load. The inclusion of a diesel generator field flash load assignnient will not change the final result, however.
• The A and B channels for the 125 VDC batteries do not have suffic!ent capacity _to last eight hours. However, if the licensee were to use the portable AC generators to provide charging to these batteries they will have sufficient capacity. This requires that the portable AC generator to be added to the list of SBO equipment and meet the criteria in Appendix B of NUMARC 87 00, except for the one hour time requirement.
• The 250 VDC buteries in Unit 1 and the dNision U 250 VDC battery in Unit 2 do not have sufficient capacity to last for eight hours. The division 1250 V batteries (in Unit 1 and 2) supports RCIC operation and division II supports HPCI operation. Since the licensee intends to use RCIC to cope during an SBO, thu'. the division 1250 VDC batteries will be examined. The licensee's data indicates that Unit i 250 VDC (ID650) is insufficient to support the operation of RCIC during the full eight hour SUO duration.

Based on the above, we conclude tnat, except for the Unit 1250 VDC (1D650) battery, all other class 1E batteries have sufficient capability or backup charging capability to support the required loads during an eight hour SB0 event. The licensee needs to:

1) Add the portable AC generator to the list of SBO equipment and meet the criteria in Appendix B of NUMARC 87 00, except for the one hour time requirement.

2) Provide a higher battery capacity for battery 1D650, or provide charging capability to the existing battery to extend its support beyond the eight hour SBO duration.

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3. Compressed Air Licensee's Submitta! The licensee stated (13) that air operated valves used for decay hev removal to cope with a SBO for (c.ur hours have sufficient backup motive sources independent of the preferred and blacked out unit's Class 1E power supply to function for four hours. No valves require i manual operation or need backup sources for operation. In its coping assessment, the licensee stated (19) that its recommended strategy is to depressurize the reactor vessel to approximately 200 ps 2 within the first hout using one SRV. Review of Licensee's Submittal Examination of the plant FSAR Sections 7.3.1.la.1 A.2 a'nd 7.7.1.12 (12) reveals that the nuclear pressure relief system at Susquehanna includes 16 pressure relief valves each operated by a pressure relief solenoid pilot air valve. Six of these valves are part of the automatic depressurization system (ADS). Each ADS valve has its own accumulator which is sizeu to provide one ADS safety / relief valve actuation at the drywell decign pressure of 45 psig or two ADS actuations at 70% drywell design pressure (31.5 psig). Therefore, these valves have sufficient back up sources of compressed air for their operation during an eight hour SDO event. 4. Effects of Loss of Ventilation Licensee's Submittal The licensee stated (13) that it has performed a plant specific analysis to determine the effects of loss of ventilation and concluded that the only dominant areas of concern (DACs) were: 18

liPCI Room 12PF RCIC Room 128'F Main Steam Tunnel 117'F The control room (which includes four relay rooms) did not exceed 120'F and was not identified as a DAC. The two relay rooms in Unit I were calculated to reach an ambient air temperature of 94'F (lower room) and 105'F (upper room) and were not identified as DACs. The Unit 2 relay rooms were assumed to be identical to those in Unit 1. In its supplemental submittal, the licensee stated (16) that the plant-specific analysis was carried out using the Compartment Temperature Transient Analysis Program (COTTAB) which was developed in house for all compartment heat up calculations. In its analysis of " dominant areas of concern," the licensee identified the Control Structure and Reactor Building (for both units). Three COTTAB calculations were performed; one for the Unit I and Unit 2 Reactor Buildings; one for the Control Structure, and one specifically for the HPCI and RCIC rooms, These calculations included the normal lighting and instrument panel heat loads, in addition to the emergency loads powered off of the station batteries. Control Room For the control room compartment, the licensee examined (16) the temperature in the cabinets which house equipment necessary during an SBO. At Susquehanna, the Control Room Complex consists of. a common control rxm to both units, and two separate relay rooms for each unit. The licensee did not analyze each individual heat load to calculate temperature rise in the control room, and as such, the control room was assumed to be at a constant 120*F throughout the SbO event. The initial temperature for the relay rooms was assumed to be 80'F which is the maximum allowable technical specification temperature. 19

Using COTTAP, the licensee calculated t17) all cabinet temperatures to be less than 180'F, except the fire protection cabinet OC650 which was calculated to be 182*F. The licensee u:ed Appendix F to NUMARC 87 00 to conclude that the equipment inside those cabinets calculated to be less than 180'F would remain operational during an EBO. As there is no equipment required for SBO in the fire protection cabinet OC650, the licensee determined the calculated temperature of 182'F to be acceptable. Reactor Building The licensee stated (16) that a compartment heat up analysis was performeu for the reactor building (excluding the HPCI and RCIC rooms) which demon:trated that no room which houses necessary equipment reaches a temperature above 130'F. HPCI and RCIC Rooms In its coping analysis, the licensee stated (19) that a COTTAB calcu;ation was performed for the HPCI and RCIC rooms. In its calculation (16), the licensee assumed both HPCI and RCIC to operate continuously for a period of 24 hours. The only electrical loads assumed were from emergency lighting and DC powered control panels. During the SBO,it was assumed that HPCI and RCIC suction were maintained only by the + CST, and therefore, the only hot pipe heat loads are the HPCI and RCIC ' f bine steam supply exhaust lines. For the bounding case, the licensee calculated a final room temperature of 126.8'F for HPCI and 116.3*F for RCIC. The licensee stated that these temperatures are less than the four hour temperature limit of 180'F presented in NUMARC S7 00 Appendix F guidelines. N addition, as part of the reactor building heat up analysis the licensee e amined a compartment in each reactor building which houses HPCI and RCIC system piping, as these compartments contain steam leak system isolation instruments which would isolate these ystems upon l a high room temperature of 167'F. The analysis showed that the HPCI 20

r i logic circuit would not be powered during SBO, and that for RCIC a maximum temperature of 1:."F would be scached during a period of 72 hours. As a result, the licensee has concluded that RCIC system isolation for this reason is not expected. t

• Batterv Rooms in its coping a:sessment, the licensee stated (19) that no specific compartment heat up analysis was performed for the battery roomo The licensee stated that the normal temperature of these rooms is between 60'F and 90*F. A lower bound for the temperature at which the battery cells begin to degrade is approximately 160*F according to IEEE 535. Since the battery cells contribute only a small amount of heat into the room, a temperature rise of 70'F to 100*F was considered to be highly unlikely.

Review of Licensee's Submittal Our review of the licensee's room heatup analyses are summarized as follows; hielbsdoloev: The COTTAP computer code that was used extensively by the licensee has not been shown tc, be adequately qualified for subcompartment analysis. The licensee has not provided any documentation on COTTAP methodology. There has been no evidence of benchmarking or quality assurance of this computer code.

• There is no evidence that any of the calculations have been reviewed or checked by the license <

Some COTTAP calculations involve a very large number of rooms which are interconnected via heat conduction pathways and airflow. No justification has been provided to substantiate if this 21

4 8 large and complex model can actually calculate a conservative and bounding 500 temperature for each room. With the information that has been provided, it is impossible to determine whether specific selection of individual room thermodynamic conditions, coupled with heat transfer paths is appropriate for every room. The calculation for the reactor building neat up includes transient temperature plots for 105 rooms that were all calculated simultaneously. These plots show different trends of temperature response with time, in some ca,es, room temperature rapidly rises within the first hour and then asymptotically approaches the steady state value on a much flatter slope. For other rooms, however, this behavior is reversed with the temperature decreasing. Still other rooms experience relatively flat temperature profiles. Finally,in some cases the slope in temperature is more extreme and may even exhibit oscillatory shaps. The licensee needs to provide a detailed technical explanation tot this variety of rcom temperature profiles. Initial Conditions: In its compartment heat up analyses, the licensee used non-conservative initial room temperatures in many cases (i.e. Switchgear rooms 93'F, HPCI/RCIC rooms 100'F). Further, the licensee assumed different and non conservative values for outside air temperature. These values range from 73.3'F to 95'F. The licensee needs to use as an initial temperature the maximum temperature allowed by technical specifications, or for the case of the outside ambient air, a bounding high summer day value. The licensee can choose a lower temperature (i.e. 78'F) as an initial temperature if it provides administrative controls to ensure that the control room temperature will not exceed this temperature under any circumstances during normal plant operation, j 22

p The licensee assumes a variety of values for ini;ial room humidity. A higher concentration of water vapor in a room will reduce the calculated room temperature. Therefore,it is conservative to use i low values of humidity. Without sufficient technical justification, the licensee must use the lowest value of humidity for each room. j Throughout the calculations, the licensee assumes a concrete thermal conductivity of 1.0 (British units). This value has previously been considered too high and therefore non-conservative for SBO analysis. A more appropriate and acceptable value of 0.7 (British units) needs to be used. The licensee needs to provide technical justification for making 180'F the maximum allowable temperature inside the control room cabinets. In its calculation of the temperature response inside the control room instrumentation cabinets the licensee assumed, as a bounding condition, that the entire control room temperature remained at 120'F for the duration of the four hour SBO event. The licensee needs to provide technical justification for the conservatism of this assumption. The licensee needs to provide justification that the COTTAP computer code calculates the maximum hot spot temperature inside the control room instrumentation cabinets, instead of the average cabinet temperature. EfluhE We were unable to verify the licensMs calculations with the information provided, at the heat loads were not cleativ mentified for each room. For example, the licensee indicated that the temperature of the Main Steam Tunnel would be 117*F after 72 hours. Upon further review of 23

i the licensee *s calculations (16), we found that the initial room temperature of the Main Steam Tunnel was assumed to be 130'F and that the temperature decreased during the course of the SB0 ivent. This indicates that either zero or very small heat generation was considered in the licensee's heat up calculation for this area. The licensee's input parameters for this area were checked and the following heat loads were identified: lighting (9,215 BTU /hr), heat removal ( 893,592 BTU /hr) and fans (72,300 BTU /hr). These heat loads will not be applicable during an SBO event. Further review of the licensee's calculation (COTTAP piping data card) did not identify any additional loads for this room. This confirms our concern that no heat loads were considered in the heat up calculation for this room. Considering the amount of heat removed from this area during normal plant operation,it is expected that the temperature of this room would rise, not decrease as indicated by the licensee, during an SBO event. This temperature rise would be expected even though the reactor would be depressurized to approximately 200 psia., as this would result in only a modest decrease in the surface temperature of the insulated main steam lines.

== Conclusion:== Based on the above, the licensee needs to provide additional information and /or technical justification for several initial conditions and modelling assumptions before we can verify the accuracy of the reported results. If the licensee cannot provide adequate justification,it may need to re-analyze the temperature response for all rooms identified as SBO dominant areas of concern. In order to better understand the calculations, the licensee needs to document the individual heat loads and assumptions for each room separately in a form that can be clearly understood. 24

5. Containment isoittion Licensee's Submittal The licensee stated (13) that the plant list of containment isolation valves (CIVs) has been reviewed to verify that valves which must be capable of being closed or that must be operated (cycled) under station blackout conditions can be positioned (with indication) Independent of the unit's preferred and Class 1E AC power supplies. The licensee added that no plant modifications and/or associated procedure changes were determined to be required to ensure that appropriate containment integrity can be provided under SBO conditions. -l I Review of Egensee's Submittal The availab!c containment isolation system data in the FSAR was examined (Table 6.2-12 and Figures 6.2 44) Upon examination of the available information and applying the containment isolation valve exclusion criteria of NUMARC 87-00 Section 7.2.5., the following valves were identified as requiring valve position indication or closure under the conditions of a SBO: ....a.........u................................. Penetratiom

== Description:== Valve Arrangement: t X 203A,B C,D PRR Pump Suetion Suppression Pool AC-GT, NO, fail as4 X.204 A,B RHR Test I.ine AC-GT, NC, fail as is X 205A,B Containment Spray , AC GT, NC, fall as is X 206A,B CS Pump Suction Suppression Pool AC-GT, NO, f ail as-is X 207A,B CS Test Suppression Pool AC CB, NC, fail as is X 208A,B CS Recircult. Son Suppression Pool AC-GT, NO, fail as is 25

In response to questions raised at the June 14,1991 telephon-conference, the licensee stated (17) that containmel t isolation can be I assured for each of the penetrations identified above by isolating all " side paths" coming off the main path downstream of the penetration. We were unable to verify that the downstream valves identified by the licensee were part of the penetration boundary or if they were in the immediate vicinity of the penetration. The use of downstream valves does not conform to the guidance. Thus, the licensee needs to list the valves identified above in an appropriate procedure and identify the actions necessary to ensure that these valves are fully closed,if needed, upon the loss of AC power. The valve closure needs to be confirmed by position indication (local, mechanical, remote, process information, etc.). 6. Reactor Coolant inventory Licensee's Submittal The licensee stated (13) that the ability to maintain adequate reactor coolant system inventory to ensure that the core is cooled has been ast,essed for four hours using a calculated leakage of 100 gpm. Review of Licensee's Submittal The licensee stated (19) that it intends to use both HPCI ard RCIC to t maintain RCS inventory during an SBO. According to the Susquehanna FSAR, the HPCI pump has a rated flow of 5070 gpm at 1172 psia reactor pressure, while the RCIC pump has a rated flow of 600 gym. It is our understanding that the RCIC system will be used when the level is established because it is easier to control than HPCI. The mjection capability of either system exceeds the amount of water required to remove decay heat and to replenish the assumed RCS leak rate of 61 gpm (18 grm per pump plus 25 gpm for maximum allowed technical specificatic,a leakage). Therefore, Susquehanna has sufficient 26 g,-,.._ ._.,-g7 y,,. - -y.,.- .,-g,_ y.,%3 ,f .w m ,-,,w--w-y ,,,y-,,.. w,. 3 .-w, ,~. c,_ ,iiw

capcbility to maintain reactor coolant inventory for the 8 hour SBO event. NOTE: The 18 gpm reactor recirculation pump sqal leak rate wrs agreed y to between NUM Ar and the NFC staff pending resolution of Generic Issue (GI) 23. If the final r:no'utiori of GI-23 defines higher RCP eal leak rates than assumed for the RCS inventory j evaluation, the licensee needs to be aware of the potential impact of this resolution on its analyses and actions addressing 4 conformance to the SBO tule. 3.3 Proposed Procedure and Training Licensee's Submittal The licensee stated (13) that plant procedures have been reviewed and modified, as necessary, to meet the guidelines of NUMARC 87-00, Section 4 in the following areas: AC pcwer restoration, (PP&I Procedure EO-000-031) per NUMARC 87-00, Section 4.2.1 and 4.2.2; ) Battery C, d5 'PP&L Procedure EO-100-030) per NUMARC 87 00 Section 4.2.1 iten 6: Severe weather, (PP&L Procedure On-000-02) per NUMARC 87-00, Section 4.2.!, Review of Licensee's Submittal We neither received nor reviewed the affected procedures, although several procedure changes have been identified as being required to maintain containre integrity under SBO condi: ions. We consider these procedures 4 to be plant specific actions concerning the required activities to cope with an SBO. It is the licensee's responsibility to revise and implement these 27 I \\

procedures, as needed, to mitigate an SBO event and to assure that these procedures are complete and correct, and that the associated training needs are carried out accordingly. 3.4 Proposed Modification Licensee's Submittal The licensee did not identify any modifications to assure a four hour coping capabil;cy as being necessary.. Ravirw of Licensee's Submittal Our evaluation found several areas where the licensee needs to perform re-evaluations, some of these may result in modifications / changes to the existing equipment. 5 Quality Assurance and Technical Specifications The licensee stated that a quality assurance program will be developed and incorpo: ited unto a plant procedure. 28

a

4.0 CONCLUSION

S Based on our review of the licensee's submittals (13,14) and the information available in the FSAR for Susquehanna Steam Electric Station Units 1 and 2 (12), we und that the submittal conforms with the requirements of the SBO rule and the guidance of RG 1.155 with the following exceptions: 1. Proposed Station Blackout Duration The licenser proposed an SBO coping duration of four hours, based on ESW group "2," an EAC classification "B," and a proposed EDG target . reliability of 0.95. Our review indicates that the Susquehanna site is ESW group "3," with an EAC classification "D," requiring an EDG target reliability of 0.975 and a minimum coping duration of eight hours. 2. Condensate Inventory Our review indicates that the CST has insufficient capacity to provide adequate core cooling for the entire eight hour coping duration. Thus, the licensee needs to provide for adequate core cooling for the entire eight hour SBO duration either by proceduralizing the manual connection of the CST to the RWST or by utilizing the cool water available in the Suppression pool at the onset of SBO to supplement or replace the CST and ensure the availability of the high pressure injection systems. 3. Class-1E Battery Capacity We conclude that, except for the Unit 1250 VDC (ID650) battery, all other class-1E batteries have suffic;ent capability or backup charging capab!Hty to support the required loads during an eight hour SBO event. The licensee needs to: l l 29

1) Add the portable AC generator to the list of SBO equipment and meet the criteria in Appendix B of NUMARC S7-00, except for the one hour time requirement.

2) Provide a higher battery capacity for battery 1D650, or provide charging capability to the existing battery to extend its support beyond the eight hour SBO duration.

4. Effects of Loss of Ventilation Our review indicates several concerns with regard to the initial conditions, modelling assumptions and res'f ts of the licensee's temperature rise calculations, as discussed in Section 3.2. The licensee needs to provide addlidonal information and /or technical justification for each concern before we can verify the accuracy of the reported results. If the licensee cannot provido adequate justification,it may need to re-analyze the temperature response for all rooms identified as SBO dominant areas of concern. In order to better understand the calculations, the licensee needs to document the individual heat loads and assumptions for each room separately in a form that can be clearly understood. 5. Containment Isolation Our review identified several containment isolation valves (CIVs) which do not meet the CIV exclusion criteria of NUMARC 87-00 Section 7.2.5. The licensee needs to list thase valves (identified in Section 3.2) in an appropriate procedure and identify the actions necessary to ensure that these valves are fully closed,if needed, upon the loss of AC power. The valve closure needs to be confirmed by position indication (local, mechanical, remote, process information, etc.). l 30 1 f

6. Quality Assurance and Tecimleal Specifications Our review has identified several areas where the licensee needs to perform re evaluations, some of which may result in modifications / changes to the existing equipment. 7. Quality Assurance and Technical Specifications The licensee's submittal does not document the conformance of the plant's SBO equipment with the guidance of RG 1.155, Appendices A, and B. 31

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5.0 REFERENCES

1. The Office of Federal Register, " Code of Federal Regulations Title 10 Part 50.63," 10 CFR 50.63, January 1,1989. 2. U.S. Nuclear Regulatory Commission, " Evaluation of Station Blackout Accidents at Nuclear Power Plants - Technical Findings Related to Unresolved Safety Issue A-44," NUREG-1032, Baranowsky, P.W., June 1988. 3. U.S. Nuclear Regulatory Commission, " Collection and Evaluation of Complew and Partial Losses of Offsite Power at Nuclear Power Plants," NUREG/CR-3992, February 1985. 4. U.S. Nuclear Regulatory Commission," Reliability of Emergency AC Power System at Nuclear Power Plants," NUREC/CR-2989, July 1983. ' 5. L S. Nuclear Regulatory Commission, " Emergency Diesel Generator Operating Experience, 1981-1983," NUREG/CR-4347, December 1985. 6. U.S Nuclear Regulatory Commission, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," NUREG/CR-3226, May 1983. - 7. U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, " Regulatory Guide 1.155 Station Blackout," August 1983. 8. Nuclear Management and Resources Council, Inc., " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Esar,ects," NUMARC 87-00, November 1987. 9. Nuclear Safety Analysis Center,"The Reliability of Emergency Diesel Generators at U.S. Nuclea: Power Plants," NSAC-108, Wyckoff, H., September 1986. 10. Thadani, A. C., Letter to W. H. Rasin of NUMARC, " Approval of NUMARC l Documents on Station Blackout (TAC-40577)," dated October 7,1988. i l 32 1 t a

f 11. Thadani, A. C., letter to A. Marion of NUMARC, " Publicly Noticed Meeting December 27, 1989," dated January 3,1990, (Confirming "NUMARC 87 00 Supplemental Questions / Answers," December 27, 1989). 12. Susquehanna Steam Electric Station Units 1 & 2, Final Safety Analysis Report (FSAR). 13. Keiser H. W., letter to Nuclear Regulatory Regulation, "Susquehanna Steam Electric Station, Station Blackout Rule," Docket No's. 50-387, 388, dated April 17,1989 14. Keiser H. W., letter to Nuclear Regulatory Regulation, "Susquehanna Steam Electric Station, Station Blackout," Docket No's. 50-387,388, dated April 17,

1990, 15.

Keiser H. W., letter to Nuclear Regulatory Regulation, "Susquehanna Steam Electric Station, Station Blackout," Docket No's. 50-387,388, dated February 27, 1991. 16. Maron, D., letter to Nuclear Regulatory Regulation, " Station Blackout Submittal and Emergency Diesel Generators," Docket No's. s0-387,388, dated May 14,1991. 17, Maron, A. K., letter to Nuclcar Regulatory Regulation, " Response to Follow-Up Questions on Susquehanna SBO Submittal," Docket No's. 50-387, 388, dated August 1,1991. 18. Russell, W. T., letter to W. Rasin of NUMARC, " Station Blackout," dated June 6,1990. 19. C. A. Boschetti," Coping Assessment for the Susquehanna Steam Electric l Station During a Station Blackcut," PP&L Technical Report No. NPE 89-04, dated April 11,1989. 20. Thom, H. C. S.,1968. "New Distributions of Extreme Winds in the United States." J. Struc. Div., ASCE 94:1787-1801. 33 l}}