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| number = ML20063P931 | | number = ML20063P931 | ||
| issue date = 07/18/1990 | | issue date = 07/18/1990 | ||
| title = Technical Evaluation Rept Prairie Island Nuclear Generating Plant Units 1 & 2 Station Blackout Evaluation | | title = Technical Evaluation Rept Prairie Island Nuclear Generating Plant Units 1 & 2 Station Blackout Evaluation | ||
| author name = | | author name = | ||
| author affiliation = SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY | | author affiliation = SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY | ||
Revision as of 16:27, 6 January 2021
| ML20063P931 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/18/1990 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20063P932 | List: |
| References | |
| CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-89-1641, TAC-68588, TAC-68589, NUDOCS 9007270156 | |
| Download: ML20063P931 (26) | |
Text
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ENCLOSURE 2
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'] IECHNICAL EVALUATION REPORT- +
i PRAIRIE !$ LAND NUCLEAR GENERATING PLANT , l; . .
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-6.. STATION BLACK 0UT EVALUATION. ;i
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( J F ,, U.S. Nuclear Regulatory Commission -! a . i. . Washington, D.C. 20555 Contract NRC-03-87-029
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Task Order No. 38 ! 1710 Goodridge Drive. PO. Box 1303. McLean, Virginia 22102 (703) 821 4300 orw sa c o~.ce, n.a.ese so u e. cwoma so..nes owan. n on.ae us seue we ane,. oa n.se, oma rwa ano se o.ree swrw ee r.asan
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- x. TABLE OF. CONTENTS Section a fagt 1.0 1 BACKGROUND ........................................... 1 2.0 REVIEW PROCESS ....................................... . 3. ,
3.0- EVAL.UATION:........................................... 6.
-3.1- Proposed Station Blackout Duration -. . . . . .-. . . .-. . . . 6 3.2 Alternate AC Power Source .......................-- 11 3.3 Station Bl ackout Coping Capability . . . . . . . . . . . . . . 13L 3.4 Proposed Procedures and Training ................ 18-
.3.5 Proposed Modifications ...............-........... -
19 m 3.6 Quality Assurance and Technical Specifications .. 20 i
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4.0 CONCLUSION
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5.0 REFERENCES
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TECHNICAL EVALUATION REPORT PRAIRIE ISLAND' NUCLEAR: GENERATING PLANT- i UNIT N0s,1 AND 2-STATION BLACK 0UT EVALUATION' 1 L 1.0- BACKGROUND' On July 21, 1988, the Nuclear Regulatory Commission-(NRC) amended its- I regulations in 10 CFR Part 50 by adding a new section, 50.63, " Loss of All Alternating Current Power" (1). The objective of this requirement is to- ' assure that all nuclear pcwer plants are capable of withstanding a station blackout (SBO)'and maintaining adequate reactor core cooling and appropriate-E containment integrity for a required duration. This requirement is' based on I information developed under the commission study of Unresolved Safety Issue A 44, " Station Blackout" (2 6). l The staff issued Regulatory Guide (RG) 1.155, " Station Blackout," to provide guidance for' meeting the requirements of 10 CFR 50.63 (7). Concurrent with the_ development of this regulatory guide, the Nuclear Utility Management and Resource rouncil (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 ! 580 rule. The NRC staff reviewed the guideline; and analysis methodology in NUMARC 87-00 and concluded that the NUMARC document provides an acceptable - guidance for addressing the 10 CFR 50.63 requirements. The application of c this method results in selecting a minimum acceptable SB0 duration capability from two to sixteen hours depending on the plant's characteristics and vulnerabilities to the risk from station blackout. The plant's . characteristics affecting the required coping capability are: the redundancy of the onsite emergency AC power so'urces, the reliability of onsite emergency power sources, the frequency of loss of offsite power (LOOP), and the probable time to restore offsite power. ' 1 4 i i
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g- , In order to achieve a consistent systema, tic response from licensees to: ' the SB0 rule.and to'e'xpedite the: staff review' process, NUMARC-developed two! ;
- generic response documents.
, These documents were reviewed and endorsed by the:
' NRCistaff (12):for the purposes of plant specific submittals. .The documents- j 1 are titled:= l
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- 1. - " Generic Response to Station Blackout Rule for Plants Using Alternate AC Power," and j
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j 2. " Generic Response to Station Blackout Rule for Plant's Using. AC. Independent Station Blackout Response Power.",. < l l- A plant-specific submittal, using one of the above generic formats, j provides'only a summary of results of the analysis of the plant's station' l blackout coping capability. Licensees are expected to ensure that the
- l baseline assumptions used .in NUMARC 87-00 are applicable to.their. plants and l to . verify the accuracy of the stated results. Compliance with. the 580 rule- ,
i 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 SB0 rule. ;. ll 4
-In'1989, a joint NRC/SAIC team headed by an NRC staff member performed.
audit. reviews of the methodology. and documentation that support the -licenseas' l submitt'als for several plants. These audits revealed several deficiencie's' l
;which were not apparent from the review of-the licensees' submittals using the j agreed upon generic response format. These deficienciesiraised a generic question.regarding the degree of the licensees' conformanceito the l
. requirements of the SB0 rule. To resolve this question, or. January 4, 1990- , ;
- NUMARC issued additional. guidance as NUMARC 87-00 Supplemental r Questions / Answers (14) addressing the NRC's concerns regarding'the l deficiencies. NUMARC requested that the licensees send their supplemental responses to the NRC addressing these-concerns by March 30, 1990... ' r, y ,
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;2.0 REVIEW PROCESS-' . j
,i ihe review of the' licensee's- submittal is focused on the following areas i
+4 consistent with the positions of RG 1.155:.
q A. Minimum acceptable SB0 duratten (Section 3.1),
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B. SB0 coping capability (Section 3.2), i y ;
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J. D. Proposed modifications (Section 3.3), and E. . Quality assurance and technical specifications for SB0 equipment (Section 3.5). '
.For the' determination of the proposed minimum acceptable SB0 duration, the following factors in the licensee's submittal- are reviewed: .a) offsite
.pewer design = characteristics, b) emergency AC power system configuration, c) determination of the emergency diesel generator (EDG) reliability consistent '
ll 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 i
=2 of RG 1.155 provides a matrix for determining'the required coping' duration. !
'For the SB0 coping capability,<the licensee's submittal is reviewed to assess'the availability, adequacy and capability of the plant systems-and ;
components needed to achieve and maintain a: safe shutdown condition and recover from an SB0 of acceptable duration which is determined above. The l review process follows the guidelines given in RG 1.155, Section 3.2, to i assure:
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- a. availability of sufficient condensate inventory .for decay heat removal, K
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b.. ' adequacy of.the clast IE battery , capacity to: support safe' ) f '
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- c. availability of adeuuate. compressed air for air operated valves necessary'for safe shutdown, ,
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- d. adequacy of the ventilation systems in the vital and/or dominant ~
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- e. ability to provide appropriate containment-integrity,.andL y,
g- :; ability of the plant to maintain adequate reactor coolant system f. m . inventory to ensure core cooling for the required coping duration. ] The licensee's submittal is reviewed to verify that requiEed procedures . f (i.e., revised existing and new) for coping with SB0 are identified and that= [ appropriate operator training will be provided. O. F The licensee's submittal for any proposed modifications to emergency. AC- i L L sources, battery capacity, condensate capacity, compressed air capacity, I L!/ -e. . appropriate containment integrity and primary coolant make up capability.is
+' reviewed. Technical specifications and quality assurance set forth- by the - 1 licensee-to ensure high reliability of- the equipment, specifically added or- -
' assigned to meet the requirements-of the SBO rule, are assessed-for their l
-adequacy.
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,l m The licensee's proposed use of.an alternate AC power source is reviewed j pf ' to determine whether it meets the criteria and guidelines of Section 3.3.5 of- ; h RG 1.155 and Appendix B of NUMARC 87-00, i l
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F s This SB0 evaluation is based on the review of the licensee's submittals. ,
'I j* dated April 13, 1989 (10) and March 29,1990(13),- and the information available in the' plant Updated Final Safety Analysis Report (UFSAR) (11); it 4
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documentation.- Suchlantaudit may be warranted'as an additional' confirmatory- i action.< This determination would be made and the~ audit would be. scheduled and-performed by the-NRC staff at some later date. -! i q 1 i
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N,7 1 . 3.0: . EVALUATION l .a, Y p>g , - $ < 1, ^ 3.1 , Proposed Station Blackout Duration + 3 . Licensee submittal '
'i The. licensee,NorthernStatePowerCompany(NSC), calculated (10and13) .1 L a minimum acceptable 580 duration of 4-hours for the Prairie Island o
Nuclear Generating Plant (PINGP) Unit Nos. I and 2 site. The licensee- j
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- stated that in order to attain this coping duration category the .)
h existing 4'16 kV station. auxiliary system will be modified to provide two dedi.cated emergency diesel generators (EDGs) for each unit. The' licensee has proposed to install two class IE EDGs independent of the existing electrical and support systems, see Section 3.5 for more j detail, d v l l The plant factors used to estimate the-proposed SB0 duration are: l 1 1. - Offsite Power Design Characteristics n. [ 3( 5 The plant AC power design characteristic group is "P2" based on: u '
- a. Independence of the plant offsite. power system s: '
characteristics of "II/2," M , W , b. Estimated frequency of LOOPS due to extremely severe weather m (ESW).which places the plant in ESW group "3," q;-u : , ps . --
- c. Expected frequency of grid related. LOOPS of less thanLone'
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- d. Estimated frequency of LOOPS due to severe weather (SW) a which places the plant in SW group "2." [
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'2. l Emergency'AC (EAC) Power Configuration-Groupi ~
= Thei EAC power configuration of the plant is. "C." Each unit atU
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PINGP will be equipped ~with'two emergency diesel generators. 40ne a EAC power supply per unit is necessary to operate' safe shutdown; ,
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equipment following a loss of offsite power. j
.3.- , Target Emergency Diesel Generator (EDG) Reliability? ,
The licensee-has selected a target EDG reliability of.0.975. The selection of this target reliability is based on having an;aserage g EDG reliability of greater than 0.95 for the last 100' demands i consistent with NUMARC 87 00, Section 3.2.4. Review of Licensee's submittal
- Factors ~which affect the estimation of the SB0 coping duration'are: the-independence of the offsite power system grouping, the estimated ,
I frequency of. LOOPS due to ESW and SW conditions, the expected frequency; of grid-related LOOPS,'the classification of EAC, and the selection of EDG. target reliability. The licensee's estimation of the-frequency,of' ) LOOPS.due .to ESW condition conforms with that given in Tables 3 2,of. 1
'NUMARC;87_00. Using the data provided in Table:3-3 of NUMARC 87 00, the- 4 expectad frequency of LOOPS at PINGS due to SW condition is estimated to i be;"0.0193".or "0.0091." This is' based on the site having offsite po'wert
; transmission lines on one, or multiple rights-of way, which. places, the
. site in SW group "3" or "2," respectively. The licensee stated .that the- 2 y
plant _is in SW group "2." Therefore, PINGP must have power transmission = , lines on two or more rights-of-way. A review of the PINGP UFSAR and the definition of the right-of-way, provided in Question / Answer No. 22 ofs " Responses to Questions Raised at NUMARC 87-00 Seminars (October,1988)', j indicate that the site could be considered as having power transmission; lines.on two rights-of way. 7 a s
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The licensee stated thattthe; independence of the; plant offsite power. ! system grouping is'"11/2." -A review of PINGP VFSAR-indicates that:-
-1. All of the.offsite power sources are connected.to two. electrically connected switchyards.
- 2. Only one of the Unit I safeguard buses,' Bus 15, can be powered directly from the unit auxiliary transformer which is fed from the
~ unit main generator. Other safeguard buses (Bus No.-16, 25, and
- 26) can only be po_wered through startup', or reserve transformers (IR and 2R transformers). Since neither the VFSAR nor the.- ,
submittal identify the normal power line-up of the emergency' -
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buses, we' assumed that.in each unit, during power operatian, all-safeguard buses are powered from the startup transformer (preferred power source). .
- 3. -The-power to these transformers is' fed from two'different power sourc e s .' The Unit I reserve transformer is fed from a 161 kV and-the Unit 2 transformer is fed from a 345 kV offsite' power source.
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s Each offsite' power source is of sufficient capacity to supply all [ critical loads for either or both units. , l,
- 4. The plant is equipped with bus tiesL(i.e. 2RYBT, .12RYBT, and , (
IRYBT) which allow crossfeeding between units.should one of- the. I start-uptransformersbeoutofservice,seeFigur_e1l(10).
- 5. Upon loss of preferred power sou'rce, (i.e loss of power from
' start up transformer), 'all emergency buses _ can'.be connected to an i
alternate power source by a manual transfer. The alternate power source for each of the emergency 4.16 kV buses is through the i corresponding emergency bus of opposing unit. This is'only ) available when the proposed modifications are completed. 8 1 _L
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6n ;The cooling tower _ substation which can be connected to a 4'.16'kV emergency in.each unit:can not be considered as an alternate power source'since it does'not conform.to~the guidance provided in Table 5 of RG 1.155. The_ guidance requires that the alternate power-l source be connectible to all safe- shutdown buses.- Using the guidance provided in Table' $ of RG 1.155 the independence of offsite power characteristics _can be categorized as "12."
' The.'EAC' classification of PINGP after implementation of the proposed-modification ~will be "C." We are unable to verify the assignment:of the' EDG. target reliability at this time._ However, based on'the information
'in_theNSAC-108,'whichpivestheEDGreliabilitydataatU.S. nuclear-reactors for calendar years 1983 to 1985, the EDGs at PINGP_ experienced-an average of 44 valid start demands per calendar year with an average-
< unit: reliability of 0.996 per diesel per year. Using this data, it appears that the target EDG reliability (0.975) selected by theilicensee; (10) to be appropriate. An audit may be requiredito ensure compliance; and to identify _ whether the licensee has 'any formal' EDG- reliability '
program consistent with the guidance of the RG 1.155, Section 1.2, and NUMARC 87-00, Appendix D. With regard to the expected: frequency of grid related LOOPS at the sih, we can not confirm the-stated-results.' The'available information in
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NUREG/CR 3992 (3), which gives a compendium ofLinformation on the loss
. of offsite power at nuclear power. plants in the U.S.. indicates- that PINGP did not have any grid-related LOOP priorito the calendar: year n 1984.
Based on above evaluations, the plant can be categorized as "P2" with <a
-minimum required SB0 coping duration of four hours.
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, The licensee: stated that, once the modifications are completed, an AAC, power souica will'be available at PINGP that meets the criteria ,
u specified in At.cwd'x B to NUMARC 87-00. = The AAC' source will~ be an EAC power sourts wtwh aeets the assumption 2.3.l(b): of NUMARC- 87-00. The'- AA7 power sorge *. Y an emergency diesel generator of the non- " f 3 blacked out wit, eeo hgure 2 (10). The AAC power source will be available wRhir )" minutes from the onset of an SB0 event and'will.have-u., n sufficient capat,ity and capability to operate systems necessary for; ', coping:with an SB0 for the required duration of 4 hours to bring and: maintain.the plant.in safe shutdown. [
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. Review of' Licensee's Submittal '
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, ' Based onLinformation provided in the plant submittal,~the AAC power source (one of the class lE EDGs) meets the criteria stated in Appendix B'of NUMARC 87 00. Each 'of the new EDGs has 5400 kW' capacity,'which is almost twice the capacity of the existing emergency diesel generator.
The existing-EDGs is rated at 3000 kW withca maximum capacity.of 3250.
.kW. The-installationofnewEDGs'meetsstrictercriteria,~(i.e. .j i'l criteria . applicable'to the class lE power supply), than th'ose provided in Appendix B to'NUMARC 87 00. The licensee needs only to demonstrate
]J by a test that the AAC power source can be established within.10 minutes- "
from the onset of an SBO, in accordance with NUMARC 87-00 Section 7.1.2 ! an RG l'155, Section 3.3.5.3. l (/ , The only concern that may aris'e from using an existing EOG as an AAC g power source at"a multi unit site where the EDGs per unit just meet the 'i
% minimum redundancy requirements is how the EDG does conform to the
[, . guidance as documented in NUMARC 87-00 Supplemental Questions / Answers J f 11 a I 7 s h f ;!.
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, 3 (14) and further explained in References 15, 16 and 17. Our review of !
tSe list of the EDG loads given in Table C.4 1 of plant UFSAR against ! T the above guidance indicates that each of the existing EDGs, which has a I smaller capacity than that of one of the new EDGs, has sufficient i g capacity to power the specified loads in the blacked out unit and the required loads in the non blacked out unit. In fact,.each of ihe l existing EDGs can support the loads required for the design basis ; accident in one and the hot shutdown in another unit with a concurrent : LOOP even't. Therefore, the AAC power source meets the requirements of ! p the 580 ruTe and the guidance provided in References 15, 16 and 17. r 3.3 Station Blar.<out coping capability . ! The licensee stated that since the AAC power source will be available within 10 minutes the coping evaluations for class IE battery capacity, compressed air, and containment isolation need not be addressed in accordance with 10 CFR 50.63(c)(2). We consider the licensee's statement to mean that ! the functions needed to cope with an SB0 are available, and that they are. adequately powered from the AAC power source for the required duration. The plant coping capability with an SB0 event for the required duration of 4 hours is assessed based on the following results:
- 1. Condensate Inventory for Decay Heat Removal t
Licensee's submittal i The licensee's submittal stated that 89,000 gallons of water are i required for the decay heat removal for four hours and for rapid cooldown of reactor coolant system (RCS) to minimize reactor coolant pump (RCP) seal, leakage. The minimum permissible condensate storage tank level per technical specification provides 100,000 gallons of water which exceeds the required quantity for coping with a 4 hour station blackout. 13
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Review of Licensee's Submittal The licensee stated that 89,000 gallons of water are needed for decay heat removal and rapid plant cooldown during a 4 hour $80 event. Since the licensee did not specify whether the 89,000 l gallons are needed for each unit or for the site, we assumed that ! it is for each unit. Using the expression provided in NUMARC 87-00, we have estimated that a total of -37,300 gallons of. condensate water would be needed to remove decay heat from the reactor core during a 4 hour SB0 event. This estimate is based on i a maximum reactor thermal power of 1683 MWt (as given in Table ; 11.9 2 of the plant FSAR for cooldown calculations), j T e cooldown process is performed by following the Westinghouse
- emergency guidelines ECA 0.0. Based on these guidelines and the information in the plant UFSAR, the plant cool down is expected to
{ be to a hot leg temperature of approximately 350*F. To' estimate ; the required condensate to cool down the RCS from an average RCS : temperature of 567'T to 350'F, we need the total storcd energy of ; the RCS (including both' metal and liquid) for each unit. Since we .! were unable to infer this infcrmation from the plant UFSAR, we , performed a bounding calculations using the RCS stored entegy data ; from a similar PWR plant. We also estimated the additional : required condensate to supplement the level shrinkage in the steam generators during the cooldown. Our calculations indicate that the condensate requirements for each unit is bounded by the licensee's estimated requirement of 89,000 gallons. I The licensee stated that the minimum condensate inventory per technical specifications is 100,000 gallons. The licensee did not specify whether this minimum condensate inventory is for each unit or for the site. According to the NRC project manager for PINGP, 14 )
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the 100,000 gallons technical specification minimum CST level is i for each unit. i ! Based on the above, we agree with the licensee that the site has i ! sufficient condensate inventory to cope and recover,from a 4-hour; 3 SB0 duration. ! l 2. Class lE Battery Capacity Licensee's submittal ; f The AAC power source will supply the battery charger (s) within 10 ; minutes of the onset of an SB0 event. Therefore no analysis of f battery capacity calculation is needed. , i Review of Licensee's submittal j i A review of the plant UFSAR indicates that each unit has two - battery banks. Each battery bank is capable of carrying the connected Icad for one hour without charge. Since the AAC power , source will be availabic within 10 minutes an'd supply the battery charger (s), we conclude tNit the plant has' sufficient battery , capacity. ; i
- 3. Compressed Air ,
Licensee's Submittal Since the AAC power source will be available within 10 minutes, no' analysis of the compressed air system is provided. 1 i 15
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v Le Review of Licensee's submittal .
.)
F l The list of the EDG loads in the plant UFSAR references to -l powering instrument air at both units after a DBA accident in one i unit concurrent with a site LOOP event. Therefore, we consider l that the instrument air is available during an SB0 event. This is j consistent with the licensee's submittal that the equipment l necessary to cope with an $80 in accordance with NUMARC 87-00, Section 7, will be powered.
- 4. Effects of Loss of. Ventilation j i
Licensee's Submittal 3 The licensee stated that the AAC power source provides powlr to heating, ventilation and air conditioning (HVAC) systems sirving dominant areas of concern to achieve and maintain safe sh9tdown - during an SB0 event. Therefore, consistent with the NUM4RC 87 00,- l Sections 7.2.4 and 7.2.1, the effects of loss of ventilation were ' not assessed. The licensee added that no modifications and/or i procedures are required to provide reasonable assurance for operability of ventilation equipment. , Review of Licensee's submittal
- The licensee's action is consistent with the guidance provided by NUMARC 87 00 and the NRC staff. We would like to emphasize that the dominant areas of concern should not be limited only to those~ '
identified in NUMARC 87 00. The licensee needs to ensure that other areas which have. heat generation sources, i.e. operating equipment, are provided with appropriate area cooling. r 16 l
.' l, a( M j F ~ . t* 5. Containment Isolation , a Licensee's submittal l The licensee stated that the AAC power source which will be d available within 10 minutes, therefore, no analysis of the J containmentisolationvalves(CIVs)isnecessary. Review of Licensee's Submittal The licensee's action is consistent with the guidance'provided in i HUMARC 87 00, Section 7.1.2. Our evaluation of the AAC power , source indicates that one division of the safety buses will be ' powered from'the AAC power source. With power available to one ) division, the licensee needs to verify that appropriate containment integrity can be assured during an SB0 event. f i
- 6. Reactor Coolant Inventory -i i
Licensee's submittal !
'The licensee stated that the AAC source powers the necessary makeup systems to maintain adequate reactor coolant system inventory-to ensure that the core is cooled for the required coping duration. ,
Review of Licensee's Submittal ; 4 , Reactor coolant makeup is necessary to replenish the RCS inventory losses due to the RCP seal leakage (25 gpm per pump per NUMARC_'87-00 guideline), the technical specifications maximum allowable leakage (estimated to be 25 gpm), and the water volume shrinkage i due to the RCS cooldown. The make up, or the charging, system at PINGP has three positive displacement pumps. Each pump has a 17
- , ,----,,,w, - - - - - , - - - - , , -- e. sw
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., design flow capacity of 60.5 gpm., If the RCS were to cool down to; e a hot leg temperature of 350'F, the RCS level shrink alone would require a makeup of -35.0 gpm.to maintain the RCS volume-(or the pressurizer level) at a level corresponding to the full load condition. With an assumed 75 gpm RCS leak rate, a rakeup system with a minimum pump capacity of 110.0 gpm would, th'crefore, be required. This translates to the requirement thsc two-out of-three charging pumps be operational from the onset of an SB0 event. However, if only one charging pump is operating, our analysis indicates that the core will not be uncovered during the 4 hour SB0 event and that the reactor cooldown is sustained by natural circulation and reflux condensation in the steam generator , U tuoes. 3.4 Proposed Procedure and Training Licensee's Submittal The licensee stated that the following plant procedures have been reviewed per guidel_ines in NUMARC 87 00, Section 4:
- 1. Station blackout response guidelines,
- 2. AC power restoration, and
- 3. Severe weather.
The licensee listed the names of the plant procedures, which fall in each of above categories, in the SB0 submittal. The licensee stated that these procedures will be, revised, if necessary, to meet NUMARC 87-00 guidelines. The licensee has identified an additional procedure change to use the non safeguards diesel-generators, (two DGs), in an interim basis until the class 'E EDGs are built. Either of these DGs has sufficient capacity to provide power for blackout loads for both 18 i
o j f units. The licensee indicated that this interim procedure does not meet the one hour time response required of an AAC sources. Review of Licensee's submittal We neither received nor reviewed the affected SB0 procedures. We consider these procedures as plant specific actions concerning the required activities to cope with an SBO. We believe that it is the licensee's responsibility to revise and implement.these procedures, as needed, to mitigate.an SB0 event and to assure that these procedures are
' complete and correct, and that the associated training needs are carried out accordingly.
3.5 Proposed Modifications Licensee Submittal The licensee stated that modifications are required to separate the existing 4.16 kV station auxiliary system into independent systems for each unit._ The existing system provides two EDGs, each of which is capable of providing EAC power to the associated emergency buses of both units, see Figure 1. The planned modification will provide two dedicated EDGs for each unit. These EDGs will be independent of the other unit's electrical and support systems and will be capable of being manually connected to the other unit's associated 4.16 kV bus using the. bus tie circuit breaker. Interlocks will be provided so that the bus tie circuit breakers must be open before the EDG output circuit breakers could be closed. After EDG output breaker closure, the bus tie breaker can be manually closed. Procedures for operation, maintenance, surveillance, and testing of the modified 4.16 kV station auxiliary system will be revised, or created as necessary. Emergency operating procedures 1EAC-0.0 and 2EAC-0.0 will be
,.2., .. . ..
>* o revised to instruct the operators to utilize the AAC power source if -
both EDGs for a unit are unavailable. Review of Licensee's Submittal j The proposed modification (see Figure 2) is. consistent with the guidance r provided in NUMARC 87 00 and RG 1.155. The installation of new EDGs' will meet stricter criteria than those given for an AAC power source, thetefore, their use as an AAC source conforms with the guidance provided in RG 1.155 and NUMARC 87 00. a ami 3.6 Quality Assurance and Technical Specifications The licensee did not provide any information concerning the conformance with the guidance of-RG 1.155, Appendices A and B. t
~
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o L * '.' 4.0 CONCLU$10NS , Based on our review of the licensee's submit 441 and the information available in the UFSAR for Prairie Island Nuclear Generating Station (PINGP), we find that'the submittal conforms with the requirements of the $80 rule and the guidance of RG 1.155 with the following exceptions:
- 1. Emergency Diesel Generator Reliability Program The licensee's submittal does not document the conformance of the plant's EDG reliability program with the guidance of the RG 1.155, Section 1.2 and NUMARC 87 00, Appendix 0, however, it is stated that the EDG target reliability of 0.975 will be maintained.
- 2. Effects of Loss of Ventilation The licentee stated that the AAC power source will be available within 10 minutes and provides power to HVAC systems serving dominant areas of concern. We would like to emphasize that the dominant areas of concern should not be limited to those identified in NUF 7C 87 00. The licensee needs to ensure that other areas whitn nause operating systems, (heat generation soutees), during an SB0 event are provided with appropriate area cooling.
- 3. Quality Assurance and Technical Specifications The licensee's submittal does not document the conformance of the plant's SB0 equipment with the guidance of RG 1.155, Appendices A, and B in its submittal.,
21 i i
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5.0 REFERENCES
l r !
- 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 I 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 ;
Complete and Partial losses of Offsite Power at Nuclear Power Plants," NUREG/CR 3992, February 1985. l 4
- 4. U.S. Nuclear Regulatory Commission, " Reliability of Emergency AC Power System at Nuclear Power Plants," NUREG/CR 2989, July 1983.
i
- 5. U.S. Nuclear Regulatory Commission, " Emergency Diesel Generator
]
Operating Cxperience, 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. l
- 7. U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, " Regulatory Guide 1.155 Station Blackout," August 1988.
- 8. Nuclear Management and Resources Council, Inc., " Guidelines and ,
Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," NUMARC 87 00, November 1987.
- 9. Nuclear Safety Analysis Center, "The Reliability of Emergency Diesel :
Generators at U.S. Nuclear Power Plants," NSAC 108, Wyckoff, H.,. " September 1986. t 22 er w *r-t-' - -- -___ *_._+___ -__ J ___-.s
o 4
+
- 10. Musolf D., letter to the Director of Dffice of Nuclear Reactor ,
Regulation U.S. Nuclear' Regulatory Commission. Attention: Document-Control Desk. " Prairie Island Nuclear Generating Plant Docket Nos. 50.282 and 50 306,. License Nos. DPR 42 and DPR 60, Loss of all alternating Current Power Information Required by 10 CFR Part-50, Section 50.63(c)(1)," dated April 13, 1989.
- 11. Prairie Island Nuclear Generating Plant, Updated Final Safety Analysis Report.
- 12. Thadani, A. C., letter with attachment to W. H. Rasin of NUMARC,
" Approval of WUMARC Documents on Station Blackout (TAC-40577)," dated Detober 7 1988.
- 13. Parker, T. M., letter to the Document Control Desk of the U.S. Nuclear Regulatory Commission, " Prairie Island Nuclear Generating Plant Docket =
Nos. 50 282 and 50 306 , License Nos. Dpr 42 and DPR 60, Supplemental' Response to Loss of All Alternating Current Power Information Required by 10 CFR 50, Section 50.63(c)(1)," dated March 29, 1990,
- 14. Thadani, A. C., letter with attachment to A. Marion of NUMARC, " Publicly Noticed Meeting December 27, 1989," dated Janusry 3, 1990 (Confirming
NUMARC 87 00 Supplemental Questions / Answers," December 27,1989).
- 15. Tam, P. S., memorandum for " Daily Highlight Forthcoming Meeting with NUMARC or Station Blackout (SBO) Issues (TAC 40577)," (Providing a Draft Staff Position Regarding Use of Emergency Ac Power Sources (EDGs) as Alternate AC (AAC) power Sources, dated April 24,1990), dated April 25, 1990.
- 16. Rosa, F., letter to Duquesne , Light Company Beaver Valley Units _1 and
- 2. " Meeting Summary Meeting of February 22, 1990, on Station Blackout issues (*iAC 68510/68511)," Docket Nos. 50 334 and 50 412, dated March 6, 1990.
23
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- 17. Russell, W. T.', letter to W, Rasin of NUMARC, " STATION BLACK 0UT
- dated l June 6, 1990. ;
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