IR 05000413/1993028
| ML20059C060 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/10/1993 |
| From: | Julian C, Shymlock M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20059C049 | List: |
| References | |
| 50-413-93-28, 50-414-93-28, NUDOCS 9401040350 | |
| Download: ML20059C060 (23) | |
Text
i e UNITED STATES
/pn nec ,% NUCLEAR REGULATORY COMMISSION y .
81 REGION ll
,
i ^% 101 MARIETTA STREET. N.W., SUITE 2900 7, j ATLANTA, GEORGIA 30323-0199
,
e
,
s
%, .....f Report Nos.: 50-413/93-28 and 50-414/93-28
'
Licensee: Duke Power Company 1
,
422. South Church Street Charlotte, NC 28242 Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52 i Facility Name: Catawba 1 and 2 i Inspection Conducted: November 1 - 5, 1993 Team Leader: fn M M. B. Sh~ymlo6K, Chief 42-3 93 Date Signed !
Plant Systems Section Engineering Branch Division of Reactor Safety Team Members: G. MacDonald, Reactor Inspector, Region II A. Pal, Electrical Systems Engineer, NRR S. Rudisail, Reactor Inspector, Region II D. Shum, Reactor Systems Engineer, NRR I Approved by: 0 i
" 1 N C. Julian, Chief Engineering Branch () Date Signed Division of Reactor Safety EXECUTIVE SUMMARY This was the first Station Blackout Inspection conducted by the Region II staff. The team used Temporary Instruction 2515/120 " Inspection of Implementatien of Station Blackout Rule Multi-Plant Action Item A-22". This inspection was to verify the adequacy of the licensee programs, procedures, training, equipment and systems, and supporting documentation for implementation of the Station Blackout (SB0) Rule,10 CFR Part 50.6 In the areas inspected no violations were identified. A deviation was identified for failure to meet the commitment identified in your letter dated of February 28, 1992. These commitments were developed based on the Safety Evaluation Report for Catawba Units 1 and 2 dated January 10, 199 PDR ADOCK 05000413 0 PDR
. _ . _ _ _
!
t i
!
!
The coping duration analysis determined that Catawba was required to cope with l an SB0 of 4-hours. The team determined that the analysis used to make this determination was performed in accordance with the guidelines established in NUMARC 87-00. The documentation to support this determination provided an -l adequate basis for the coping duration. The team concluded that the station ;
blackout systems were adequate to cope with a station blackout of a 4-hour .
duratio ,
v The station blackout procedures provided operators with adequate instructions .
to operate station blackout equipment during an SB0 event. Procedural steps !
for the determination of SB0 conditions were adequate and for the declaration ;
of an SB0 they were judged to be good. Procedural controls to maintain !
Reactor Coolant System inventory were adequate to prevent the core from being uncovered during the four hour coping period. The team concluded that the emergency lighting and communications equipment were adequate to successfully operate SB0 equipment and coordinate emergency operating procedure activities !
during an SB0 event. Station Blackout training provided to the operating i staff was thorough and included simulator scenario training with an SB ;
In the area of electrical calculations, the team considered the licensee's !
response to be poor. Calculations were either incomplete or had not been performed. However, the licensee responded to the team in this area by ;
providing calculations that were performed during the inspection. Several calculations were revised and some computer codes rerun to provide the bases i for the team to assess the adequacy of the specific design. At the conclusion of the inspection the licensee outlined the calculations that would be prepared or revised to complete the documentation to support implementation of-the requirements of the SB0 rul The team found the testing of the SB0 related equipment to be adequate. The l reliability programs for the Emergency Diesel Generators and the Safe Shutdown !
Facility diesel were determined to be adequate. The team also determined that !
the quality assurance program currently specified for the SB0 equipment met i the requirements of Regulatory Guide 1.155 Appendix !
'
Overall, the team considered the procedures, training, and equipment to be adequate for the SB0 program implementation. However, the team considered the !
documentation, specifically in the area of calculations and the response to i the recommendation of the safety evaluation report to be wea !
,
I f
l i
I l
I
!
'
, -
- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
TABLE OF CONTENTS EXECUTIVE SUMMARY 1.0 Inspection Background /0bjective .................. I Safety Evaluation Recommendations . . . . . . . . . . . . . . . . . 1 Coping Duration Analysis ..................... 3 St ati on Bl ackout Systems . . . . . . . . . . . . . . . . . . . . . . 3 Alternate Alternating Current (AAC) Power Source ...... 3 Battery Systems . . . . . . . . . . . . . . . . . . . . . . . 5 Auxiliary Feedwater And Steam Relief ............ 6 Condensate Inventory .................... 7 Effects of the Loss of Ventilation ........ .... 7 Containment Isolation . . . . . . . . . . . . . . . . . . . . 9 Compressed Air ....................... 9 Reactor Cool ant Inventory . . . . . . . . . . . . . . . . . . 9 Diesel Generator Reliability Program ............ 10 4.10 Chemical and Volume Control . . . . . . . . . . . . . . . . . 11 4.11 Emergency Lighting and Communications . . . . . . . . . . . . 11 4.12 Heat Tracing ........................ 12 4.13 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . 12 4.14 Station Blackout Electrical Calculations .......... 13 Station Bl ackout Procedures . . . . . . . . . . . . . . . . . . . . 14
! Emergency Response Procedures . . . . . . . . . . . . . . . . 14 Severe Weather Procedures . . . . . . . . . . . . . . . . . . 15 Station Blackout Training . . . . . . . . . . . . . . .. . . . . . . 16 I i Station Blackout Equipment Quality Assurance Program ....... 16 i
! Exit Meeting ........................... 18 Appendix A - Persons Contacted Appendix B - Acronyms And Abbreviations l
. .
- - - - - - - - -
s r
F 1.0 Inspection Background /0bjective I In 1988, the NRC issued the Station Blackout Rule,10 CFR Part 50.63,
" Loss of All Alternating Current Power." Guidance on acceptable methods ;
for meeting the requirements of the rule were established in NRC regulatory guide 1.155, " Station Blackout." Concurrent with the development of the regulatory guide, the Nuclear Management and Resource Council (NUMARC) developed guidelines and procedures for assessing station coping capability and duration. This was documented in NUMARC 87-00, " Guidelines and Technical Bases for NUMARC Initiatives Addressing ;
Station Blackout at Light Water Reactors."
This inspection was the first station blackout inspection conducted in Regirn II. The inspection was conducted using Temporary Instruction 2515/120, Inspection of Implementation of Station Blackout Rul ' Safety Evaluation Recommendations
'
The team reviewed the implementation of the recommendations contained in the Catawba SB0 Supplemental Safety Evaluation (SSE). The licensee had implemented three of the six recommendations outlined in the Safety '
Evaluation (SE). By letter dated February 28, 1992, the licensee indicated that Catawba would implement the recommendations of the SB0 SE by December 31, 1992. Catawba did not meet this implementation dat Failure to implement these commitments is identified as Deviation 50-413,414/93-28-0 The Catawba SB0 SE contained the following recommendations: "The licensee needs to provide assurance (1) that the Class IE ,
vital I&C battery loads that could occur during an SB0 event would not exceed those measured during the blackout test, (2) ensure that the batteries have sufficient capacity for closing of the ;
necessary circuit breakers to restore offsite power, and (3) ,
ensure the EDG batteries have sufficient capacity for EDG field fl ashi ng. "
The licensee had not completed the implementation of this recommendatio Refer to section 4.2 of this report for detail . "The licensee should ensure the accessibility to the steam generator PORV's and auxiliary feedwater flow control valves and the habitability in the areas where these valves are located during an SB0 event."
The licensee had not completed the implementation of this recommendation. The team reviewed SB0 emergency response ;
'
procedure EP/1,2/A/5000/03, Loss Of All AC Power. The procedure called for local manual operation of AFW flow control valves to !
control S/G water level. Calculation CNC-1211.00-00-Olll, i Mechanical Penetration Area Temperature Analysis, dated March 23, 1993, determined the temperature of the mechanical penetration i room in which AFW flow control valves were located. The '
i
.
.
!
,
!
.
calculated peak temperature of the mechanical penetration room was approximately 175 *F. During the inspection, the licensee revised the calculation and removed the heat sources which would not be present under SB0 conditions. The revised maximum temperature for the mechanical penetration room was approximately 122*F. This ,
calculation was performed for actual plant conditions. The team reviewed the revised calculation and determined that it was '
acceptable. The team concluded that recommendations 2 and 5 had been resolved and implemented during the inspectio . "The licensee should verify that power will be available for the turbine driven auxiliary feedwater (TDAFW) pump pit ventilation fan during an SB0 event."
The team verified by review of station drawings and walkdowns that the TDAFW pump pit ventilation fan would be powered by the SSF diesel generator (DG).
4. "The licensee should provide a procedure which will require the operators to open instrument cabinet doors within 30 minutes following an SB0 in accordance with the guidance described in NUMARC 87-00." ,
The team reviewed procedure EP/1,2/A/5000/03, Loss Of All AC Power, and verified that the procedure contained instructions for opening instrument cabinet doors within 30 minutes of the initiation of SB . "The licensee should verify that no manual operation of SB0 response equipment in the annulus and mechanical penetration rooms is required during an SB0 event. The licensee should also verify that the calculation for the McGuire mechanical penetration room ;
is applicable to Catawba."
The team verified, by review of station drawings and the '
procedure,EP/1,2/A/5000/03, Loss of All AC Power, that the mechanical penetration rooms were the only DACs where local manual operation of equipment would be required during an SB0 event. For
'
the accessibility and habitability in the mechanical rooms refer to the discussion of recommendation 2 abov . "The licensee should provide confirmation and include in the documentation supporting the SB0 submittals that a program meeting as a minimum the guidance of RG 1.155, position 1.2, is in place or will be implemented."
The licensee had completed this recommendation. The licensee's emergency power source reliability program is discussed in sections 4.1 and 4.9 of this repor t i
e
,
3.0 Coping Duration Analysis
The Station Blackout Rule required that systems provide sufficient .*
capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a .
'
station blackout for the specified duration. The specified duration was calculated based on factors such as the off-site power design ;
characteristics, emergency alternating current (AC) power supply !
configuration, and the emergency diesel generator reliability. A ,
minimum acceptable SB0 duration of four hours was calculated for the !
Catawba Nuclear Station Units 1 and 2. The team verified selected factors which were used to determine the station blackout duratio The licensee provided the proposed 4-hour station blackout duration in a letter to the NRC dated April 17, 1989. The NRC reviewed the proposed SB0 duration and agreed with the licensee's evaluation as documented in >
the NRC Safety Evaluation Report dated January 10, 1992. The four hour station blackout duration was based on an off-site power design ,
characteristic group of "P1," an emergency ac configuration group of '
"C," and an emergency diesel generator (EDG) reliability target of 0.9 The off-site power design characteristic group "Pl" was derived from an !
independence of off-site power characteristic of "I2", severe weather group "1", extremely severe weather group of "1", and an expected loss of off-site power of 1 ss than 1 per 20 years. The team reviewed plant documentation to verify that these characteristics were appropriat The team reviewed the EDG reliability data used to calculate the EDG unit average reliability values. The reliability values (unit average)
for the last 100 valid start and load run demands were 0.95S for Unit 1 and 0.985 for Unit 2. The reliability data for the last 50 valid start and load run demands were 0.98 for Unit I and 0.97 for Unit 2. The i reliability data for the last 20 valid start and load run demands were 1.0 for Units 1 and 2. Based on a review of the diesel reliability data, the team determined that the selection of a 0.95 EDG reliability '
target value was aopropriat The team concluded that the calculated minimum acceptable station '
blackout duration of four hours was in accordance with the NUMARC guidelines and was acceptable.
4.0 Station Blackout Systems 4.1 Alternate Alternating Current (AAC) Power Source The licensee committed to provide an AAC power source to the selected SB0 loads by utilizing the Appendix R dedicated safe shutdown facility (SSF) diesel generator (DG) within 10 minutes of the onset of an SB ,
The team reviewed the AAC power source to verify that this AAC power source met the criteria specified in NUMARC 87-00, Appendix , ,,n - - - - - - - -
.. _, _ _ . _ _ _ _ _ _ _ _ - - . _
I-l j
i
i
!
The Appendix R SSF DG was rated at 600 V, 875 KVA (700 KW), at a power i factor of 0.8. The SSF DG does not provide power to all the SB0 load In response to a' concern raised by the team regarding the protection of f the AAC System against likely weather related events, the licensee f indicated that the SSF DG was located in a structure which conforms to .l Uniform Building Codes and necessary cabling between buildings was .l buried.- The licensee further stated that the wind load of 30 pounds per j square foot (PSF) or a wind speed of 95 mph and live load of-50 PSF -
(including snow and ice) was used as design requirements for the SS These ratings were consistent with those used for Category I structure
~
design !
i The licensee provided SSF DG loading calculation (CNC-1381.06-00-0052, j Rev.14 dated 11/2/93 ). The team identified the following comments- -J i
1) The assumed power factor (pf) of 0.9 and the efficiency of'O.9 wa !
non-conservativ .l
2) The basis for the load factor (0.5) used was not supporte j!
3) Cable loss and transformer loss were not considere ) Maximum Transient load was not calculate ) KVA and HP for individual loads were added algebraicall i 6) The calculation should include KW and KVAR of each load and tota {
KW, KVAR and KV i
!
The team noted that the licensee had not performed a voltage drop j calculation for the SB0 loads when' fed from the SSF DG. A preliminary ;
calculation was performed on 11/4/93 to verify that adequate voltage- !
will be available _ at the SB0 equipment terminals. The team had'several j comments about that calculatio :
l 1) The source voltage in the calculation was 600 V rather than 540 SSF DG testing indicated acceptable voltage of 600+/- 60 volt ) Nameplate data or test values were not used for the load data u (locked rotor current, locked rotor pf, etc.) !
3) Cable temperature was not based on the maximum room temperature during an SB0 event.
_
4) The effect of SB0 temperature on the equipment was not considere ) Low voltage (120V or less) equipment terminal voltage was not calculate l The licensee indicated that they would revise the SSF DG loading calculation and finalize the voltage drop calculation incorporating the team comment .- -. .
- - - . -
.
.
I' ,
The team further noted that SSF DG had 0 failures in the last 20 valid start and load run tests, 2 failures in the last 50 valid start and load run tests and 2 failures in the last 100 valid start and load run test Based on the above, the SSF DG had a reliability of greater than 0.9 l The team noted that the SSF DG Reliability Program was similar to the EDG Reliability Progra The SSF System has been tested monthly per !
l Technical Specifications (TS) 4.7.13. However, TS does not include an 7 acceptance criteria for voltage and frequency of the SSF D Per NUMARC 87-00, criteria B.10, once every refueling outage, a timed start and rated load capacity test shall be performed. The licensee '
,
indicated that monthly testing covers this test. The team noted that ;
'
timed start (time to start the SSF DG and connect to the safety bus within 10 minutes) was not covered in the testing procedures. However, -
the licensee's procedure (Loss of All AC Power - Procedure N ,
EP/1/A/5000/03 Revision 12) required reactor coolant pump seal cooling !
within 10 minute !
The licensee verified that the AAC (SSF DG) power source could be started and provide power to the safety bus within 10-minutes. Based on this requirement the team found this acceptabl The team concluded that the SSF DG had sufficient capacity to power selected SB0 loads, and had a reliability of greater than 0.95 and met all the requirements of NUMARC 87-00, Appendix .2 Battery Systems
'!
The team reviewed the vital battery capacity calculation (CNC-1381.05- l 00-0122, Rev.13 dated 11/1/93). This calculation was performed to '
assure that the battery capacity was adequate to provide power to required SB0 equipment for four hour During the review of the calculation, the team verified that the battery j sizing calculation included an appropriate aging factor of 1.25, and I temperature correction factor of 1.048 (69.16*F electrolyte l temperature). The team found that the input current requirements for j the constant KW loads such as the Uninteruptible Power Supply were adjusted properly using the lowest voltage of 105 V. However, the team identified the following weaknesses in the calculation:
1) The SE Recommendation of the last minute load of breaker closing was not considere i 2) The temperature correction factor was not based on the lowest ;
electrolyte temperature of 60*F per the FSAR or consistent with .
other battery calculation temperature I The team found that calculations to demonstrate that adequate voltage was available at the equipment terminals fed from vital battery had not ;
been performed. Other calculations were reviewed. The team reviewed the SSF Battery Sizing calculation and found it to be acceptabl __ .
i
The team found that the following calculations for the battery systems at Catawba which are required to support an SB0 had not been performed: ;
'
1) SSF Battery Voltage Calculation 2) EDG Battery Sizing Calculation 3) EDG Battery Voltage Calculation
,
The team concluded that the licensee deviated from their commitment of completing all supporting calculations by December 31, 1992. However, the licensee indicated that they would complete all necessary calculations by June, 199 :
4.3 Auxiliary Feedwater And Steam Relief Steam Generators (S/G) B and C and their associated Auxiliary Feedwater (AFW) flow control valves and power operated relief valves (PORVs) are i utilized for decay heat removal during an SB0 event. The AFW flow control valves and the PORVs are located in the mechanical penetration .
rooms and the inboard doghouses, respectively. The AFW flow control valves are air operated and fail closed on loss of air. During an SB0 event the AFW flow control valves are manually operated. The S/G PORVs i have a nitrogen backup system and can be operated remotely from the control room. If nitrogen pressure were lost the valves can be operated locall The team verified that the SB0 procedures had detailed instructions to locally operate these AFW flow control valves and S/G PORVs. Training on the manual operation of these valves was provided to the plant operator The team also verified that communication and emergency lighting equipment was available to operate these valves during an SB0 even The team reviewed heat-up calculations CNC-1211.00-00-0111, Mechanical Penetration Area Temperature Analysis, dated March 23, 1993 and CNC-1240.00-00-0006, SSF Temperature Calculation / Auxiliary Feedwater Pump Area, dated May 4,1984, to determine if the above areas were accessible and habitable during an SB0 event. The calculated peak temperatures in the mechanical penetration rooms where the AFW flow control valves are located was 122.l*F. These valves would be operated on an intermittent basis during an SB0 even The team concluded that the procedures and training provided to the operators for operating the auxiliary feedwater and steam relief systems during an SB0 event were appropriate. The AFW flow control valves were accessible to plant operators during an SB0 event and temperatures in the vicinity of the valves would not prevent local operation of these valve >
- - _ . .
.
. 7 Condensate Inventory The team reviewed licensee calculation CNC-1240.02-00-0001, Rev.11, NUMARC 87-00 SB0 Mechanical Compliance, to determine if the Catawba plant had adequate condensate to cope with an SB0 event of 4-hour dur'atio The results of the condensate inventory calculation which was performed in accordance with the guidance provided in NUMARC 87-00, Section 7.2.1, indicated that 75,452 gallons of condensate were required for decay heat -
removal during an SB0 event of 4-hour duration. The turbine-driven auxiliary feedwater pump (TDAFWP), during normal operations was aligned '
to the AFW condensate storage tank (45,000 gallons, shared between units), the upper surge tank (85,000 gallons) and the condenser hotwell (170,000 gallons). During an SB0 event this alignment would be the '
same. The SB0 procedures. contained guidance to isolate the AFW condensate storage tank and the upper surge tank upon their depletio !
In addition to the condensate sources described above, procedure AP/1,2/A/5500/06, change 0, retype 12," Loss Of S/G Feedwater", also had instructions to align the TDAFWP suction to the condenser circulating :
water system which had the capacity for approximately 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of reactor coolant system decay heat removal and cooldown during an SB0 7 event. The licensee provided training to plant operators on the i isolation of the above tanks and the transfer of the TDAFWP suction from these tanks to the condenser circulating water syste .
The team concluded that the condensate / water inventory at the Catawba ,
plant was adequate to cope with an SB0 event of a 4-hour duratio ; Effects of the Loss of Ventilation ,
As indicated in the NRC Safety Evaluatica, dated January 10, 1992, pertaining to the licensee's initial response to the SB0 rule, the licensee identified the containment, annulus, AFW pump room, TDAFWP pit, ,
mechanical penetration. rooms and inboard doghouses as dominant areas of concern (DACs). The team reviewed the following heat-up rate calculations; CNC-1211.00-00-0111, dated March 23, 1993, Mechanical Penetration Area Temperature Analysis; CNC-1240.00-00-006, dated May 4, 1984, SSF Temperature Calculation / Auxiliary Feedwater Pump Area; CNC-1211.00-00-0102, dated April 3,1992, Battery Cubicle-Maximum and i Minimum Temperatures; CNC-1211.00-00-0031, dated August 28, 1984, ,
Containment Temperature Analysis During Loss of AC Power. These t calculations identified the DACs. The team reviewed the following calculations which evaluated operability of SB0 mitigation equipment l located in the DACs; CNC-1210.04-00-0052, dated April 19,1989, Appendix !
R Qualification of SSS Components; CNC-1381.05-00-0118, Rev.1, Catawba Nuclear Station Unit I and 2 Station Blackout Coping Study. The SB0 t procedures that provide compensatory measures taken for the loss the ,
ventilation systems during a station blackout event were also reviewed i by the tea !
.
The team verified that the heat generation rate calculations for various equipment, piping and components in the areas containing equipment required to cope with an 5B0 event were conservatively performed and that the peak temperatures in these areas (except for the containment)
during a 4-hour SB0 event were properly calculated. All the calculated ,
peak area temperatures during a 4-hour 580 event were less than the temperature limits for various equipment operabilities as described in NUMARC 87-00. Therefore, the reasonable assurance of operability for equipment required to cope with an SSO event was acceptable, in calculating the peak temperature in the containment, the containment ,
was assumed to be one volume and the temperature in the entire containment volume was at a uniform value. The calculated peak temperature was 205'F. The team found that the containment was not properly modeled in the calculation.
The containment consists of three major compartments (upper and lower compartments and the ice condenser compartment). The lower compartment volume is approximately 1/3 of the total containment volume. With the exception of a small allowable ice condenser bypass flow area, there is ,
no communication (flow path) between the upper and lower compartments during an SB0 event. During an SB0 event, most of the heat sources will ;
be in the lower compartment. The team requested that the licensee rerun ,
the containment temperature calculation using a computer code to properly model the containment compartments and the heat sinks. The re-calculated peak containment temperature was 255'F which was well below the calculated peak temperature of 327*F resulting from a main steam line break. The team found that the re-calculated peak containment temperature of 255' F was acceptable.
In addition, the licensee used the transient heat transfer method and re-calculated the peak temperature in the mechanical penetration rooms.
In the re-calculation, the licensee used the heat generation rates based on SB0 heat sources and per the NUMARC 87-00 guidance and assumed the adjacent auxiliary building temperature to be 120*F. The re-calculated '
.
peak temperature in the mechanical penetration rooms d wing a-4-hour SB0 event is 122.l* The team verified that the heat generation rates and the peak room temperature were properly calculated and that the mechanical penetration' rooms were the only DACs where local manual operation of equipment would ba required during an SB0 event.
In the heat-up calculation for the AFW pump room and TDAPM oit, the licensee assumed that a ventilation fan would be operating during an SB0 event. The team verified that power would be available for this ventilation fan from the SSF DG during an SB0 event.
The team also verified that the SB0 procedures had provisions which required opening of control room cabinet doors within 30 minutes after ,
the onset of an SB0 even l
,
F 9 l The team concluded that the heat-up calculations to identify the DACs !
are technically sound, the equipment located in the DACs had a reasonable assurance of operability and the mechanical penetration rooms would be accessible and habitable. In addition, the team determined that the SB0 procedures had sufficient provisions to mitigate the ,
consequences of loss of ventilation during an SB0 even '
s 4.6 Containment Isolation The team verified that the licensee had correctly identified the i containment isolation valves which did not conform with the guidance described in NUtiARC 87-00 for exclusion from consideration as isolation valves of concern. Four valves (NV918, WL8078, WL869B and WL8278) .
required manual closure and/or position verification locally during an SB0 event. These valves are accessible to operators during SB0 condition These four valves were identified in the SB0 procedure. The operators !
have been trained to verify that containment isolation (if needed) has ;
been successfully accomplished. The team concluded that the actions to provide containment integrity during an SB0 event are appropriat , Compressed Air The AFW flow control valves which are air operated during normal power i operation would be manually operated to control the S/G water level during an SB0 event. The S/G PORVs have nitrogen backup pressure from independent nitrogen supplies for remote operation and can also be manually operated during an SB0 event. Each of these PORVs have two supply bottles of nitrogen which could supply pressure for approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of PORV operation during an SB0 even The team concluded that adequate nitrogen would be available to perform the required PORV operation during an SB0 event and that no equipment ,
which would be required to cope with an SB0 event would be dependent on compressed air during an SB0 event.
4.8 Reactor Coolant Inventory l The team verified by review of the licensee's evaluations, procedure !
and drawings that the reactor coolant inventory was adequate to ensure that the reactor core was not uncovered during the four hour SB0 coping <
period. The licensee's design incorporated a standby makeup pump which would be powered from the SSF DG. This ensured that even during SB0 conditions the plant would have' a limited makeup capability-to the reactor coolant syste The licensee evaluated the reactor coolant inventory without the use of the standby makeup pump. The evaluation was performed using the Modular Accident Analysis Program (MAAP). The team reviewed the MAAP evaluation for both a cooldown case and a non-cooldown case. The evaluation i indicated that the reactor core would not be uncovered in either cas .
The team reviewed the primary assumptions in the evaluation and determined that the analysis was adequate. The analysis assumed a leak :
rate of 25 gpm per reactor coolant pump and an allowed leakage of 11 gpm. This equates to 111 gpm leak rate. The analyzed results indicated that greater than seven hours would elapse before the core would be uncovere However, the licensee has a makeup pump sized to maintain RCS inventor !
The normal seal leakage of 12 gpm (3 gpm per pump) and normal system t leakage of 13 gpm yields a total expected ieakage of 25 gpm. The makeup pump is rated far 26 gpm capacity at a head of at least 2488 psig. The ;
SB0 emergency response procedure called for the operators to start the i SSF DG and the makcup pump within 10 minutes of the SBO. These actions would take place in the SSF which is remote to the main control roo The licensee has conducted training on these evolutions. The team -
reviewed the results of the training and timing of the four shift crews and noted that all crews completed the evolution in less than 10 minute The team reviewed the TS and ver :fied that the makeup pump was being tested every 31 day >
The fuel storage pool supplies borated water to the makeup pump. The TS controlled the baron concentration and available water volume ensuring that an adequate water supply would be available to the makeup pum The TS also contained a requirement that the makeup pump be declared inoperable if the total of identified and unidentified leakage exceeds 26 gpm. The team performed a walkdown of the standby makeup pump system and verified the pump ratings and system configuration. The team reviewed procedure PT/1/A/4200/07C, Standby Makeup Pump Performance Test, and verified that the acceptance criteria met the TS requirement *
The team concluded that the procedures, controls and evaluations of the RCS inventory were adequate to prevent the core from being uncovered ;
during the four hour coping perio i 4.9 Diesel Generator Reliability Programs The team reviewed the emergency diesel generator reliability program to verify that the EDG reliability data was being trended and the program was consistent with the guidance of RG 1.155, Section '
The team noted that the EDG reliability target level consistent with the ,
plant category and coping duration _ selected was not included in the EDG Reliability program. The team found that the surveillance testing of -
EDGs was per TS 4.8.1. The Operations Management Procedure 2-28 monitored each start and load run of the EDG and trended the reliabilit Failure rates are trended. The Systems Engineering Supervisor or designee keeps track of the number of valid failures in the last 100, 50, and 20 valid tests (on a per EDG and per unit basis)
and notifies the Operation Unit Manager or designee to adjust the EDG ,
test frequency in accordance with TS. Each valid test, invalid failure, ,
or valid failure was reviewed and a Problem Investigation Process (PIP)
report was issued and recorded in the EDG logbook and the surveillance interval' would be adjusted accordingly. All EDG failures, whether valid
!
,
I
.
or invalid, would be reported to the Regulatory Compliance Section via '
the PIP. The team noted that the program was adequate in identifying responsibilities for the major program elements. Management oversight programs were adequate.for ensuring reliability levels were achieved and ensuring the reliability program was functioning properl '
The EDG failure data since 1/89 was as follows:
EDG NO. of FAILURES NO. of VALID START AND LOAD RUNS .
IA 0 90 '
18 0 86 2A 3 102 i
+
2B 1 101 t
The team concluded that the EDG reliability program was detailed and was *
consistent with the guidance of RG 1.155, Section 1.2, with the following exceptions: ;
1) Identification of an EDG Reliability target level consistent with I the plant category and coping duratio ) EDG loading criteria (Section 6.1.c of procedure 2-28 and 1.10) of ,
greater than or equal to 2875 KW for valid start and load was not ;
'
consistent with the TS requirement of greater than or equal to 5600 KW but less than or equal to 5750 K ;
The licensee indicated ~ that they would revise the applicable procedures to incorporate the above concerns.
4.10 Chemical and Volume Control System The team reviewed the RCS makeup capability to determine if the valves l in the makeup flow path would be operable independent of the preferred and SB0 unit's normal Class IE AC power source. The standby makeup pump ;
and associated valves were capable of being aligned to the SSF {
independent power supplies. The water supply for the makeup pump is '
from the fuel storage pool via a line from the fuel transfer tube. The team walked down the makeup pump and verified that the equipment could j be operated from the SSF control panel. Station drawings were reviewed '
and the switchgear and motor control centers were inspected to verify ;
that the makeup pump and associated valves were powered from SSF power source i 4.11 Emergency Lighting and Communications The team reviewed the emergency lighting and communication equipment to ,
assure that they were adequate and available to support operations i personnel during an SB0 even '
l
,
!
DC lighting was fed from the 250 V Plant Battery and would be available for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The battery charger for the 250 V Battery would not be powered from the AAC power source. Additionally, 8-hour ;
battery packs (Appendix R lighting) and portable lighting would also be available during the SB0 even '
The team verified that adequate communications would be available during an SB0 even Regular telephone, sound powered phones and 2-way radios were available. A separate diesel generator with 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> capacity was available as the backup power source for the regular telephones syste ,
The team concluded that the emergency lighting and communications l equipment were adequate to successfully operate SB0 equipment and coordinate emergency operating precedure activities during an SB0 event.
4.12 Heat Tracing NUMARC 87-00 section 4.3.1 (13) required that the licensee consider the impact on safe shutdown due to the loss of heat tracing systems. The licensee had not previously evaluated the impact of the loss of heat tracing systems on achieving and maintaining safe shutdown. The licensee performed an evaluation and determined that the heat tracing systems were not required to achieve and maintain safe shutdown under SB0 conditions for the required four hour coping duratio The only heat tracing which was significant to SB0 equipment was the heat tracing on the steam lines to the AFW pump turbine. This line was heat traced to eliminate steam condensation in the line. Under SB0 ,
ccnditions, the AFW system initiates rapidly upon loss of feedwater "
signal and there was not enough time for condensation to occur in the steam line. Once steam flow was established, condensation would not occur, and the system would continue to operate for the entire coping duration. The team reviewed the licensee's evaluation and concluded that loss of heat tracing systems would not impact the ability to achieve and maintain safe shutdown for the required coping duratio ;
4.13 Fire Protection NUMARC 87-00 section 4.3.1 (10) required that the licensee consider the impact on safe shutdown of the potential actuation of fire protection systems due to elevated temperatures under SB0 conditions. The licensee had not previously evaluated the impact of potential fire protection system actuation. In response to questions from the team, the licensee reviewed the impact of fire protection actuation on required SB0 ,
equipmen The licensee indicated that interaction between fire I detection / protection systems and equipment required during SB0 had been ,
evaluated under the IPEEE program. All systems within safety related '
areas had been reviewed for inadvertent actuation and the effects of i
f
,
t i
water spray on the surrounding environment with no concerns identifie ;
Specific areas related to SB0 equipment were the AFW pump room and the l EDG rooms.
The AFW pump rooms have carbon dioxide systems for the AFW pump pit The detectors in the AFW pump pits were set at 190*F and the calculated
- max room temperature in the area was 160*F. The licensee indicated that l the AFW pump turbines were not oxygen dependent. The team concluded
. that the fire protection systems would not impact AFW pump operation i under SB0 condition *
I The EDG rooms utilized a carbon dioxide fire protection system. The EDG room lowest temperature detectors were set at 190*F. The EDG rooms were not determined to be DAC in the SB0 loss of ventilation evaluation. The licensee had tested the EDGs and verified that they would start and run following a carbon dioxide discharge. The team concluded that the fire protection systems would not impact the ability to achieve and maintain safe shutdown following a SB0 for the required coping duratio l 4.14 Station Blackout Electrical Calculations !
During the riew of the calculations to support the implementation of the SRC rt.'+ it was determined that the required calculations were i ei ther incouplete or had not been performed. The team considered the licensee response to be poor in this area especially since there were no - !
modifications needed to meet the rule. However, the licensee responded to the team in this area by providing calculations that were performed ;
during the inspection. Several calculations were revised and some i computer codes rerun to provide the bases for the team to assess the j adequacy of the specific design. The following table indicates the l status of the calculations as understood by the team at the end of the inspection. The table also indicates the teams understanding of the licensee proposed actions that were discussed at the exi _
CALCULATION / STATUS LICENSEE PLANS TO PERFORM Class IE Battery Sizing /not complete Revise by June 1994 Class IE Battery Voltage /no Perform new calculation by June calculatio SSF Battery Sizing / complete Nothing needed SSF Battery Voltage /no calculation Perform new calculation by June '94 EDG Battery Sizing / not complete Revise by June 1994 EDG Battery Voltage /no calculation Perform new calculation by June '94 SSF Diesel Loading /not complete Revise by June 1994 ,
.SSF Diesel Voltage /no calculation Perform new calculation by June '94 j
.
!
5.0 Station Blackout Procedures 5.1 Emergency Response Procedures The team reviewed procedures and validated selected procedure steps through walkthroughs, to verify that the procedures provided instructions to mitigate an SB0 and that they were consistent with NUMARC 87-00 section The main SB0 emergency response procedure was EP/1,2/A/5000/03, " Loss Of All AC Power". This procedure was the instruction followed for SB0 immediate actions, it referenced other instructions for recovery actions, long term operation of the SSF facility, and transferring between different AFW pump condensate sources.
.
Procedure EP/1,2/A/5000/03 immediate actions provided guidance to minimize RCS inventory loss, ensure an RCS heat sink, and ensure rapid restoration of AC power via the SSF DG. The parameters for ,
determination of SB0 conditions were adequate and the declaration of SB0 !
was judged to be good. The procedure led the operator to a rapid decision that a S'] had occurred and actions were immediately prescribed to start the SSF JG and makeup pump to achieve RCS pump seal flow within 10 minutes. This portion of the procedure was determined to be goo The team verified that the procedure contained guidance to ensure that the RCS was isolated to minimize inventory loss. Procedural guidance was provided for verifying adequate AFW flow and RCS heat remova The :
procedure contained a requirement to open control room cabinet doors within 30 minutes if control room ventilation was lost. This implemented a specific NRC SSE recommendatio The team noted that the procedure provided adequate guidance regarding ;
containment isolation. A listing of CIVs was provided and manual actions were specified if required. The team verified that the CIV status panel would have power under SB0 conditions which would facilitate verification of containment isolatio The procedure required local operator action to control AFW flo Adequate guidance was provided in the procedure for this activity. The
,
emergency procedure referenced procedure AP/1,2/A/5500/06," Loss Of S/G 4 Feedwater", for guidance regarding AFW pump suction water supplies. The l team performed a procedure walkthrough of AP/1,2/A/5500/06 and i determined that instructions were provided to monitor and transfer !
condensate sources to ensure adequate condensate inventory to the TDAFW !
pump for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SB0 coping duratio ;
Instructions were contained in the emergency procedure for the operators to strip DC loads from the non-vital DC batteries to preserve battery '
capacity. The procedures also contained instructions for ensuring TDAFW ,
sump pumps were operated. The procedure contained guidance regarding i instrument air and a provision for starting a temporary compressor to I restore instrument ai !
!
.___ ______ _ _ _ - _ _ _ - .
.
. .
The team verified that the procedure contained guidance on operating the S/G PORVs. Under SB0 conditions, the S/G PORVs would be operated remotely from the control room using backup nitrogen supplies. The team verified that the S/G PORVs were also capable of local manual operation and that adequate guidance was provided in the procedures for this mode of operation. The team concluded that the procedures contained adequate guidance for local manual oper. ~, ion of the S/G PORVs and that the lighting, communication, and h itability were adequate. The S/G PORVs were each equipped with 2 nit' on bottlo In the event of loss of nitrogen, additional nitrogen ,ttles could be installe The procedures did not explicitly address portable lighting or security impacts under SB0 conditions. The team had two comments regarding enclosure 3 to EP/1,2/A/5000/03, " Establishing NC Pump Seal Injection From lhe SSF". The first comment was that step "e" of the procedure requiring adjustment of the SSF DG output frequency did not specify a tolerance in the frequency acceptance criteria. The second comment was that step 1.a appeared to be an action to be taken in the event the SSF DG failed to start and not an action the operator should verify prior to making their initial attempt to start the engine. Under the 10 minute requirement to establish makeup flow to the RCS pump seals, this activity could be performed after the initial start attempt.
5.2 Severe Weather Procedures The team reviewed the severe weather procedure to verify that the procedure provided adequate guidance to prepare the site for ensuing ,
severe weathe The Catawba site has an extremely severe weather characteristics of 1, as determined by the licensee and accepted by the NRC. This characteristics is indicative of a location with a small probability of ,
having winds in excess of 125 miles per hour. These high winds are !
normally associated with hurricanes. However, due to the passage of i Hurricane Hugo, Catawba Nuclear Station implemented a procedure to prepare the site for a hurricane. The site also implemented a procedure to prepare for tornado Procedure,RP/0/A/5000/07, Change No. 7, "
Natural Disaster and Earthquake" provides instructions for the identification and elimination of potential missiles from the site; reviewing the adequacy of site staff to support operations and repair; expediting the restoration of important plant systems and components to service and increasing CST inventory. The procedure also addresses the need for the plant to be shutdown two hours prior to arrival of the hurrican The team determined that the severe weather procedure is consistent with the guidance provided in NUMARC 87-0 . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
.
16 Station Blackout Training The team reviewed the licensee's training records to determine if the licensee had conducted adequate training to ensure that its staff could cope with an SB0 of four hour duratio The licensee training documentation identified training requirements for each specific job position. Both procedure training and specific task training were specified by job position. The team reviewed the training records for the C shift personnel. All C shift personnel had completed their SB0 training requirements except for one Non-licensed operator who was relatively new on shift and had not yet completed one of the required SB0 training module '
The team determined that the training requirements specified for SB0 were thorough and included task training required for SB0 mitigatio The team considered the training records to be adequat .
The team performed some procedure walkthroughs for certain manual actions. During the walkthroughs, the team questioned the operations personnel to determine if they were familiar with SB0 mitigation equipment location and operation. The operations personnel were ;
familiar with SB0 equipment location and operation. The team noted that there was some difficulty in locating the AFW manual stop valves in the doghouse. These valves are not used for AFW flow control during an SB0 but could be used if the mechanical penetration room could not be entere The licensee walked the team through an SB0 scenario on the plant simulator. The team reviewed the exercise for an elapsed run time of approximately one hour. The team noted that the licensee's procedures rapidly led the operations personnel to recognize the event and declare that an SB0 had occurre The AFW system automatically initiated and the team noted that no S/G high level alarm was received for approximately 45 minutes with the AFW i flow control valves fully open. Thus no manual action would be required for at least 45 minute The operations staff verified containment isolation and operated the S/G PORVs from the control room. The team considered that the scenario demonstrated that plant operations personnel were familiar with SB0 procedures and understood how to cope with an SBO. The team concluded that the licensee's SB0 procedures and training were adequate to cope with an SB0 of four hour duration. Station Blackout Equipment Quality Assurance Program -
i The team reviewed the quality assurance program applied to the SB0 ,
equipment to verify that it met the requirements outlined in RG 1.155 '
Appendix '
i
!
l l
_ _ _ _ - - -
.
,
The licensee prepared a document entitled Standby Shutdown System (SSS)
which provided guidance on identification, procurement, and design of the SSS for Catawba and McGuire. The SSS equipment was components that participated during safe shutdown. The document listed the major SSS components and identified those that were classified as Quality Level Approximately 50% of the components were designated as QA Level I components. The remaining components listed which were not Quality level I were equipment associated with the standby shutdown facility diesel and support equipment. This equipment was installed to meet 10 CFR 50 Appendix R requirements. The criteria document was intended to describe the site programs, processes, and procedures applied to the SSS equipment to ensure that SSS equipment would meet its intended design functions. The present criteria document only identified the requirements for receipt inspection, identification of SSS equipment on design drawings, and for listing the SSS equipment in the Nuclear Policy Manual / Site Directives (NSD). The criteria document should clearly describe how all the quality requirements in RG 1.155 Appendix A were met. The licensee indicated that they would revise the criteria i document to clearly describe all the quality requirements applicable to the SSS equipment. This document would be included in NSD 30 The team reviewed other site program documents to verify that the quality requirements not described in NSD 307 were applicable to the SSS equipmen NSD 208.4 Rev. I and NSD 212 Rev. O were reviewed and the team verified that the Corrective Action and Root Cause Analysis Programs were !
applicable to SB0 equipment. NSD 700 Rev. I was reviewed and the team i verified that the verification process applied to SB0 equipment. The licensee indicated that to date no audits had been performed on SB0 equipmen l
!
The team selected 2 procurement design specifications for review. CNS-1320.21-00-0004 dated March 14, 1979 on the Makeup pump was reviewe CNS-1358.04-00-0001 on the SSF Inverters was reviewed. Both specifications were prepared to include operational, environmental and technical requirements. The team judged the specifications to be goo The team determined that the quality assurance program currently specified for the SB0 Equipment met the requirements of RG 1.155 '
Appendix A. The QA program for the SSS equipment which was required for SB0 mitigation met the requirements of eithcr 10 CFR 50 Appendix B for ,
the QA level I components or 10CFR50 Appendix R for the non QA level 1 {
components. In accordance with RG 1.155 these QA programs met the i requirements specified in Appendix A of RG 1.155. The teams review of selected Nuclear Stations Directives and selected procurement specifications verified that the licensee had properly implemented QA programs for the SB0 equipmen _ _ _ - _ _ _ - _ - _
. - . . - ...
.
. -18
'
8,0 Exit Meeting The inspection scope and findings were-summarized on November 5, 1993,
'
with those persons indicated in Appendix A. The team leader described the areas inspected and discussed in detail the inspection result Dissenting comments were not received from the licensee. Proprietary-information is not contained in this repor :
Item Number Status Description 413,414/93-28-01 Open Deviation: Failure to Meet Commitment for Implementation of the SB0 rul !
l
.
>
?
,
a
I l
,
t
,
<
, - , , - - - - ----r- . __ -_-____.__m
.
Appendix A i
Persons Contacted Licensee Employees ,
- S. Brown Acting Mechanical Nuclear Engineering Manager
- T. Crawford Systems Engineering Manager
- R. Dickard Electrical Engineering Supervisor i'
- J. Forbes Engineering Manager
- E. Fritz Systems Engineering
- J. Glasser Electrical Engineer
- G. Kent Mechanical Engineer
- G. Maddox Associate Engineer j
- W. McCollum Station Manager
- K. Nicolson Compliance Specialist
- D. Rehn Site Vice-President t
- R. Smith Design Engineer i
- Z. Taylor Compliance Manager
- Thomas Electrical Engineering Manager 1 Other licensee employees contacted during this inspection included engineers, operators, technicians and administrative personne NRC Personnel
- C. Berlinger Chief, Electrical Engineering Branch, NRR
- R. Freudenberger Senior Resident Inspector
- Attended exit interview l
l i
-
i l
l
. . . ,
.
. Appendix B Acronyms and Abbreviations AC Alternating Current AAC Aiternate Alternating Current AFW Auxiliary Feedwater CFR Code of Federal Regulations CIV Containment Isolation Valve ,
DG Diesel Generator :
.DAC Dominant Area of Concern i EDG Emergency Diesel Generator i-FSAR Final Safety Analysis Report GPM Gallons per Minute KVA Kilo-volt Amperes KVAR Kilo-volt Amperes Reactive KW Kilowatts LOCA Loss of Coolant Accident MAAP Modular accident Analy is Program NUMARC Nuclear Management Resources Council PF Power Factor l PORV Power Operated Relief Valve i PSF Pounds per Square Foot PSIG Pounds per Square Inch Gauge QA Quality Assurance ,
RG Regulatory Guide _j RCS Reactor Coolant System SSF Standby Shutdown Facility SSS Standby Shutdown System SB0 Station Blackout TS Technical Specifications TDAFW Turbine Driven Auxiliary.Feedwater
20
_ _