ML20058K973

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Technical Evaluation Rept - Millstone Nuclear Power Station Unit 1,Station Blackout Evaluation, Final Rept
ML20058K973
Person / Time
Site: Millstone 
Issue date: 07/18/1990
From:
SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:
NRC
Shared Package
ML20058K974 List:
References
CON-FIN-D-1311, CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-89-1154, TAC-68566, NUDOCS 9008030210
Download: ML20058K973 (36)
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g k 5 I - [ , y'fi (; lt i y_ r + 'I .l i -[' .,.e-1 ~,. ~l w i4 j t 4 y TECHNICAL EVALUATION RFPORT w MILLSTONE NUCLEAR' POWER STATION, UNITENo. 1- - i y. w 4 Nie STATION BLACK 0UT EVALUATION a y 3, p 1 m, l9' TAC No. 68566. + .v ' Y - ~ a 't N j l i lh Final 4 July -18,- 1990 i t i 'l f -k'._ j< a 4 - i- .x Prepared,for: lg U.S. Nuclear Regulatory Commission ' Washington, D.C. 20555: contract NRC-03-87-029 Task Order No. 38 [L - 900803o2 2 3gg pcp m . Poet OtRce Bon 1323,1710 Goo &cke Drive, McLm Vwginia 22:02,1703) 821M s l 0 'I ... i.

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TABLE 0F CONTENTS

[ 5ection Paga 11'0 BACKGROUND.'.......................................... 1 2.0 REVIEW PROCESS-....................................... 3 3.0-EVALUATION'..;.;...................................... 6 3.1 Proposed-Station Blackout Duration'.............. 6 3.2 Alternate AC Power Source....................... 9 3.3 Station Blackout Coping Capability.............. 13 so 3.4 Proposed' Procedures and Training................ 25 3.5 Proposed Modifications.......................... 26 L. 3.6-Quality Assurance and Technical Specifications.. 26 ~ a 'j+

4.0 CONCLUSION

S.......................................... 28 s

5.0 REFERENCES

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j{if $,,* P / J,,. u ? m J %, 1 ~ TECHNICAL' EVALUATION REPORT-p E, -NILLSTONE NUCLEAR POWER STATION, UNIT No. 1 q. $TATION BLACK 0UT EVALUATION-l'. 0 : BACKGROUND ,g On July 21, 1988, the Nuclear Regulatory Commission (NRC) amended its regulations in 10 CFR Part 50 by adding a new section, 50.63, " Loss of All ' Alternating Current Power"1(1). The objective of this requirement is to a'ssure that' all nuclear power plants are capable of withstanding a station blackout (SBO) and maintaining adequate reactor core cooling and appropriate containment integrity for a required duration. This requirement is based on information developed under'the commission study of Unresolved Safety Issue A-44,l" Station Blackout" (2-6). 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 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 SB0 rule. The NRC staff reviewed the guidelines and analysis methodology in d

'NUMARC 8600 and concluded that,the NUMARC document provides an acceptable. N guidance for addressingsthe 10 CFR 50.63 requirements. The application of-this method results in selecting,a minimum acceptable SB0 duration capability from two to: sixteen hours depending on the plant's characteristics and , vulnerabilities to the risk from station blackout. The plant's - characteristics affecting the required coping capability are: the redundancy of the onsite emergency AC power sources, the reliability of onsite emergency power sources, the frequency of loss of offsite. power (LOOP), and the probable time,to restore offsite power. 1 1 i

^ ~ y % w jd, ; O y L.; 7.,, ~ 1 qq 1 y !!n order,to-achie've a consistent systematic response from licensees to % 'the SB0 rule Land to expedite the 'st'aff review process, NUMARC developed two [,

generic response documents..These documents were reviewed and endorsed by the; i

1> .NRC staff (17) for the purposes of plant specific submittals. The documents are titled:- Y 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 Lc blackout coping capability. Licensees are expected to-ensure that the ,1

baseliie asst.mptions used in NUMARC 87-00 are applicable to their plants and-to verify the accuracy of the. stated results.

Compliance.with the SB0 rule ~ l 7 F requit 9ments'is verified by review and evaluation of the licensee's submittal L E and audit review of.the supporting documents as necessary. Follow up NRC g g inspections assure that the licensee has implemented the necessary changes as required to meet the SB0 rule, t

In 1989, a joint NRC/SAIC team headed by an NRC staff' member. performed j

audit. reviews of the methodology and-documentation'that' support =the licensees' i 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 .i question regarding the degree of_ licensees' conformance to the requirements of the SB0 rule. To' resolve this question, on-January 4,1990, NUMARC issued f additional guidance as NUMARC 87-00 Supplemental Questions / Answers (16) j addressing.the NRC's concerns regarding the deficiencies. NUMARC requested ~ that the licensees send their supplemental -responses to the NRC addressing i 'these concerns by March 30, 1990. 2 .I4 .v.,m m ., ~-

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t 2.0 REVIEW PROCESS a

The review of theilicensee's submittal ~ is focused on the following areas-consistent with the positions of RG 1.155: A. ' Minimum acceptable SB0 duration (Section 3.1), B. S80 coping capability (Section 3.2), f t .C. Procedures and training for SB0 (Section'3.4), .D.- Proposed modifications (Section 3.3), and j E. Quality assurance and technt:a1 specifications for SBO equipment [ .(Sr tion 3.5). L 3 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) l determination of the emergency diesel generator.-(EDG) reliability-consistent q Ewith NSAC-108 criteria,(9),'and d) determination of the accepted EDG target l I. reliability. Once these factors are known,-Table 3-8 of NUMARC 87-00* or Table 2 of RG I.155 provides a matr.ix for' determining theirequired; coping. duration.- For the SBO? coping capability, the licensee's' submittal.is' reviewed to. k L - 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 il determined-above.. TheJ review process. follows the guidelines given in RG l' 155, Section 3.24 to ' assure:- n a. availability of sufficient condensate inventory for decay heat-m.m

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adequacy of_the class'lE battery capacity to support safe-3

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I c. availability of adequate compressed air for air-operated valves necessary for safe. shutdown, ~ 4 _d. adequacy of the ventilation systems-in the vital and/or dominant. areas -that include equipment necessary for safe shutdown of the

plant, e.

ability to provide appropriate containment integrity, and-i p F f. ability o.f the plant to maintain adequate reactor coolant system a u inventory to ensure core cooling for the required coping duration. r [ ? -The licensee's submittal is reviewed to verify that required procedures (i.e., revised existing and new) for coping with SB0 are identified and that ~ appropriate operator training will be.prov ded. E The licensee's submittal for'any proposed modifications to emergency AC' sources, battery capacity, condensate' capacity, compressed air' capacity, L

appropriatelcontainment integrity'and primary coolant make-up capability i's

'i reviewedL Technical specifications and-quality assurance-set forth by the lj, 1licensee to ensure high reliability of the equipment, specifically added or ? ! assigned t'o' meet' the requirements of the SB0 rule, are assessed for their adequacy. [ j-The licensee's: proposed use of an alternate AC power source is reviewed ' lto determine whether'it meets the criteria and guidelines of Section 3.3.5 of RG:1.155 and Appendix B of NUMARC 87-00.. x 1g A normal' SB0 review;is limited to the review of the licensee submittal; itidoes not include a concurrent' site audit review of the supporting X documentation. Such an audit may be warranted as an additional confirmatory n 5 4 E A lb2 i

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action. This determination would be made and the audit would be scheduledTand performed by the-NRC staff.at some later date.

9 However, a limited number of concurrent site audit reviews were performed'in order to obtain-a benchmark for licensee conformance-with the' l documentation requirements of the SB0 rule. - Millstone Nuclear Power Station ~ Unit No. I was one;of the plants selected by the NRC staff for a concurrent ~ udit review of SB0 supporting l documentation. This audit was performed by the a -joint NRC/SAIC team, headed by an NRC staff member, on July 18-21, 1989. The following evaluation was written in coordination ~with NRC staff-and encompasses both the ' review of the licensee's submittals dated April'17,1989 (10), May 30, 1989 (15), and March 30, 1990 (20), the site audit. review, a telephone conversation between the NRC/SAIC staff and the licensee on i September 7, 1989, and a follow-up response from the licensee. 5 s 4 L p.c s t s i f (. t. [ p 1-, n k -^: f' ; 'z. k o.

e MN ') ~ s. 3.0t. EVALUATION j 3.1 ' Proposed Station Blackout Duration 1 L' Licensee's submittal c L iThe licensee, Northeast Utilities (NU),' calculated (10, 15, and 20) a. jf minimum ~ acceptable SB0 duration of 8 hours for the Millstone Nuclear i p Power Station Unit No. 1 (MP1). The licensee stated that no j modification is necessary.to attain this proposed coping duration. The-plant factors used to estimate the proposed SB0 duration are: { o 1. Offsite. Power Design Characteristics o The plant AC power design characteristic group is.'P3" based on: J s a. Independence of the. plant offsite power system characteristics of "I3," ] Li I b.- Expected frequency of grid-related LOOPS of less than one j per 20 years, '5 c. Estimated' frequency of LOOPS dub io extremely severe' weather. ~ of 1.20E 02 per' year (10) which places the' plant in ESW 1 group "5,'" and o d.. Estimated frequency of LOOPS due to severe weather of ~ -1.47E-01 per year which places the plant in.SW group "5." i 2. . Emergency AC (EAC) Power Configuration Group The licensee stated that the EAC power configuration of the plant L is'"C." MP1 is equipped with two emergency power sources; one I i l' 6 i I L l l 1

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a_ 'i >'o .e g ip ' diesel; generator (DG)and'oneGasTurbine' Generator (GTG).both-j g,[ 2..,. are normally-available to the unit safe shutdown equipment. jpM i h 4 _0ne EAC power supply is necessary to operate safe shutdown - p' equipment:following a LOOP. ] o '43 i h wj// 3.. Target Emergency AC Reliability E %.

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The licensee has selected a target EAC reliability of 0.975' based. l -f1-f p on having a nuclear unit average reliability for the 20 demands greater than 0.90 consistent with NUMARC 87-00 selectionL t 'N criterion = { e y The licensee has stated that this target reliability is' justified. J by the recent improvements in GTG reliability resulting from a-g series of modifications that have-been or are planned to'be 1 imp.lemented.~ These are: a. The original carbon steel piping'and fittings.in the GTG air- [ L, start system are being replaced with stainless steel. This L Lreplacement is scheduled for completion during the 1989-refueling outage. This modification should reduce the failures of air start motors due to-rust particles.: 4

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\\ b. The GTG' air start: system was modified to provide automatic closure of the air shutoff valve when the GTG fails to i L start. This modification was done to increase GTG start' ~ t reliability-by minimizing the accumulation:of starting air- . contaminants in the regulator valve. r h c.- The GTG_ start logic will >be modified during the 1989 refuelin'g outage to bypass certain emergency protective . trips that are not presently bypassed during emergency operation. In addition, the high lube oil temperature trip 4

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. will-be modified to a coincident 'two-out-of-two logic scheme. l d. The existing electronic devices in the gas turbine' governor system will be replaced with' state-of-the-art' microprocessors during 1989 refueling outage. j j Review of Licensee's submittal 1 Our review concurs with the evaluation of th'e proposed SB0 coping duration. Factors which affect the estimation of the SB0 coping f duration are; the independence of the offsite power system grouping, the estimated frequency of LOOPS due to-the. severe and extremely severe weather conditions, the classification of EAC, and'the selection of EAC-target reliability. These= factors were found to be properly evaluated. I The licensee's estimation of the-frequency of LOOPS due to ESW and SW was based on the data v ovided in NUMARC 87-00. The independence of [ H offsite power system grouping at MP1 is "I3." All the offsite power sources at MP1 are connected to the unit's safe shutdown buses-through one switchyard. The normal source of AC power to these buses is from - the unit main generator.and there is one automatic transfer and no manual' transfer. of all safe shutdown buses.;to one preferred orione / alternate offsite_ power' source.1 _The assignment.of the EAC target reliability is supported by the demonstrated 97.5% start and load-run. -unit average reliability over the last.20 demands (11).- The statistics-provided by the licensee to justify the above reliability targetLare-(12): j-GAS TURBINE GENERATOR DIESEL GENERATOR F A I L U P E to F A.I L U R E to Start' Load-Run Start Load-Run' I Last 20 0 1 0 0-Last 50 2 1 0 0~ L last 100 4 5 0 1* [m

  • Note:

Licensee's coping evaluation identifies a total of three start and load-run failures for DG (11). 8 4 m

The above statistics indicate that the DG satisfies the conditions for a reliability target of 0.975. In the case of the GTG, the 0.975 reliability target is acceptable but it requires continued vigilance to assure high reliability. The data shows a long term GTG performance problem is being corrected, but the evidence is not strong enough to assume that it will not recur. The licensee stated that the present EDG/GTG reliability (maintenance / surveillance) program is sufficient to satisfy the guidelines of RG 1.155, Section 1.2 and NUMARC 87-00, Appendix D. The licensee stated, however, that a formal reliability program will be established in accordance with the above guidance and the targeted EAC reliability will be maintained (20). 3.2 Alternate AC (AAC) power source Licensee's Submittal The AAC power source is a previously installed alternate power supply provided to satisfy Appendix R safe shutdown capability. The AAC power supply for MP1 is from the Millstone Point Unit 2 (MP2) emergency power source through a cross tie between the two units, see Figure 1 (10). The cross tie aligns the MP1 emergency buses through a non-class IE bus, Bus 14H, to the MP2 emergency bus 24E. This allows MP1 to be powered by either of the MP2 EDGs. The licensee stated that the AAC power source is available within one hour of the onset of an SB0 event and has sufficient capacity and capability to operate systems necessary for coping with an SB0 for the required duration of eight hours to achieve and maintain the plant in a safe shutdown condition. The AAC power source meets the criteria specified in Appendix B to NUMARC 87-00 and the assumptions stated in Section 2.3.1 of NUMARC 87-00. I

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a -i L 1 . c: (.. q .o n ReviewofLiconsee'ssubmitial'- s lp i ? Our review concurs with the licensee's. conclusion except for-the following t'hree' cases concerning the AAC's conformance to the Appendix B. of NUMARC'87-00: -{ ? 4 .o

Paragraph B.3 of Appendix B of NUMARC 87-00. states,-

j " Components and subsystems shall be protected against the l L

. effects of likely weather-related events that may-initiate the loss'of offsite power event. Protection may be'provided N-

by' enclosing AAC components within structures that conform t with the Uniform Building Code, and burying exposed cable-run;between buildings." q o. Paragraph B.9 of Appendix B of NUMARC 87-00 states,."the AAC~ ' power system shall-be... capable of. maintaining voltage and - 1 frequency.within limits consistent with established indust'ry standards thet will not degrade the performance of any-i L shutdown system or component. -At multi-unit sites...., an d adjacent unit's Class IE power may be used as an AAC power source for the blacked out, unit if it is-capable of powering .4 the required loads for both units."- Paragraph B.12 of Appendix 'B of NUMARC.87-00/ states, "Unless o.- otherwise governed by technical' specifications,-the AAC' ' system shall;be demonstrated b'y initial test to be capable ? of powering required shutdown equipment within one hour of-a'- 1 station blackout event." 1 1 The '14H bus, which the licensee is proposing to use for powering the MP1 j ~ emergency buses from the MP2 emergency EDGs, is located in a cubicle in; i <+ - the yard south of the MP1 maintenance warehouse. The location of this 1 f cubicle makes the bus vulnerable to conditions (i.e. salt spray, flood, 11 d

e c s. ' hurricane,- and tornado)Lthat could cause a total loss of offsite power' l 'at the site. 'Even-if the cubicle co'uld structurally surviverthe h' f extremely severe weather conditions, it is prone to' fail electrically. Water can enter the cubicle from seepage through.the walls or from-operators entering and leaving the cubicle. In addition, the power cables cross-connecting between the units are partially exposed to the 'same conditions as those of the cubicle. The review team is also concerned with the structural integrity of both the cubicle and the power cable supports. To comply with the paragraph B.3 it is.necessary to provide protection for AAC components by enclosing the 14H bus in a structure and burying the power cables between the buildings. o-Paragraph B.9 requires that the licensee to demonstrate the adequacy of i the proposed AAC by an initial test. This test must realistically; -represent or envelope SB0 loading. conditions, including. the most severe' i expected transient as well as steady-state loading.- The specific j concern is the load transient that would occur when the Unit 1 and. ~ N F , Unit 2 4160 V circuit breakers are closed to. supply power.to the MPI's-i control rod drive pump, while the AAC power source is supporti6g other-t steady state.SBO. loads, in addition, Paragraph B.9 requires'that the: y adjacent unit EDGs, which just meet the minimum redundancy requirements J for that unit, have sufficient capabity to power all the-required loads-4 7 at both units. y We reviewed theLlist of loads on Table-8.3-2 of the MP2 FSAR, andTthose L, _ provided by the licensee during the site audit review against the NRC's 9 y guidance.regarding the.use of existing EDGs as.an AAC power source _which is documented in the SB0 rule, 'NUMARC 87-00 Supplemental Questions and m H Answers under question B.3 and 3.4 (18), and further explained in References'21, 22, and 23. Our review of the-MP2 EDG loads after an LOOP indicates that each EDG would have 900 kW of excess capacity that could be used as an AAC power source for the MP1 SB0 loads. The i licensee's estimated SB0 load is 348 kW. Therefore, we agree with the licensee that MP2 EDGs can be used as an AAC power source for MPl. L i 12 p y

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( 9 ' .R j To comply with the parag-sph B.12, the licensee needs to perform a single test demonstrating that actions required for powering necesstry shutdom equipment can be a. comp 11shed within one hour from the onset of an $B0 event. 3.3 Station Blackout Coping Capability The plant coping capability with an 580 for the required duration of L eight hours is assessed based on the following results: ) 1. Condensate Inventory for Decay Heat Removal Licencee's submittal-l The licensee's submittal stated that a total of 132,g00 gallons of water are required for decay heat removal, reactor cooldown (tr,- ) .350'F) and reactor coolant system make up for the proposed SB0 duration of eight hours. The calculated condentate water required for the decay heat and cooldown is 103,357 gallons. The minimum L permissible condensate level from two fire water tanks per j l technical specifications provides a usable volume of 378,000 vallons which exceeds the required quantity for coping with an 8 hour 180. MP1 usti on-Isolation Condenser (IC) cooled by fire water using a diese; driven fhe ;: ump to remove decay heat from the reactor core during an SB0 event. The licensee stated that no plant modifications or procedural changes are necessary to use the fire water and the IC. Review of Licensee's Submittal y The required condensate (fire water) inventory for decay heat removal (IC system) was calculated by the licensee using an in-house computer code, THIST, wMeh estimates the integrated decay 13 m_,,___________2 m

l' heat release during an 8 hour SB0 duration. The licensee stated that this computer code has been qualified for safety related calculations. For the purpose of this review we used data from Branch Technical Position ASB g 2, " Residual Decay Heat Release Rate for Light Water Reactors" (13) to estimate an eight hour integrated decay heat similar to the licensee's calculated value. The cooldown to approximately 350'F was calculated by the licensee with an analysis of the stored energy of all the water, fuel, and metal in the reactor vessel. The stored energy and cooldown calculations were reviewed and found to be acceptable. We agree with tle licensee that the plant has an adequate condensate inventory for an 8 hour SB0 duration. In addition, the available excess fire water tank inventory can be used,to assist in SB0 recovery. 2. Class lE Battery Capacity Licensee's'Submitta) The class lE station batteries ("A" and "B") at MP1 must function for only one-hour following an onset of an SBO. After the initial one hour, the AAC power source which power the battery charger for the "B" station battery in conjunction with the "A" station battery will supply the power needed for maintaining the plant in a safe shutdown condition. The licensee stated that the clast IE batteries are sized to support the safe shutdown loads for two hours without charge. Review of Licensee's Submittal The licensee's calculations of class lE batteries indicates that the "A" battery has enough capacity with a minimum load (56 amperes from 60 to 479 minutes) to be operable during an 8 hour 14 l I

\\ i $80 without charging. The 'B' battery, which supplies vital AC, computer, and emerg e.y lighting, has sufficient capacity for one hour after which in aarger is powered from the AAC power source. The licensee's calculations of the battery loading during an $80 is based on the actual required current (ammeter readings) instead of the nemeplate ratings. We believe that the use.,f the actual ammeter readings conforms to the guidance for present capacity calculations provided that: 1. The ammeter readings represent the maximum values teken over a period of testing and not just from a one time test. 2. .The licensee re evaluate the battery capacity if any change to the present plant DC loading occurs. 3. The normal battery backed plant monitoring and electrical ~ system controls in the control room remain operational during an SB0 event. These are considered to~be essential for successful coping with and recovery from an SB0 as documented in NUMARC 87-00 Supplemental Questions / Answers. 3. Compressed Air Licensee's submittal The licensee stated that no air operated valves are relied upon to cope with an SB0 for eight hours. The IC valves, which open automatically on sustained high reactor pressure, are powered from-class IE batteries and the reactor vess31 safety / relief valves have sufficient backup compressed air (nitrogen) for their i operations. 4 15 l

y j s l 4 Review of Licensee's submittal The only air operated valves required to cope with an SB0 event i = aretheautomaticdepressurizationsystemvalves(ADSV). Each ADSV, in addition to being connected to the drywell nitrogen header (14),hasanaccumulatorwithsufficientnitrogencapacity for several cycles of valve operation. The accumulator nitrogen capacity is judged to be adequate for expected ADV operations in j support'of reactor depressurization during an SB0 event. l L 4. Effects of Lo.ts of VenH 5ation Licensee's submittal l i The calculated post-SB0 8 hour steady state ambient air temperature for the plant areas containing SB0 ergipment is as ~ i follows(15): Arg.t ,Tennerature IoF) f.inal Initial Control Room 91 73 Steam Tunnel 173 104 Reactor Building (Elevation 82'-9") 168 104 Turbine Building Switchgear Area 118 134 Southeast Corner Room <86 (NUSCOtest p memo GMB 86 595) The licensee stated that only those areas having a final ambient Ie temperature greater than 120'F are considered to be dominant areas H

.of concern (DAC).

Therefore, this eliminates all rooms but the steam tunnei and reactor building (elevation 82'9"),as DACs. In the case of the steam tunnel, the licensee stated that once the 4 inboard and outboard containment isolation valves are closed, the 16 l- --y. m m.-.-,+%+- -w.~e,.+-p-ee3 -e.- --..y-gm,,-g wm.w.y

+ t i heat load in the steam tunnel would drop considerably.

However, more importantly, the licensee asserted that none of the equipment l

in the steam tunnel is needed for an SBC event because of the reliance of this plant on the IC inside the reactor building. Based on these two factors, the licensee established that the steam tunnel is not a DAC.- The reactor building (elevation 82'9") is the location of the IC and does have a calculated temperature that would make it a DAC. However, the operability of-the equipment in this room has been evaluated for a harsher environment resulting from a postulated high energy line break-(HELB) as part of the environmental equipment qualification program. Since the IC system is remotely operated from the control room, no operator entry into this area is necessary. Therefore, the licensee asserts that this area is not a DAC. The licensee stated (15 and 16) that the following procedure changes are necessary to provide reasonable assurance for. ~ equipment operability: L 1. Open the following control room cabinet panel access. doors within 30 minutes of an SB0 (16): CRP 918, CRP'925, CRP 980 2, CRP 980 4, PAM 103, PAM 104, i 2. Monitor the average bulk drywell temperature and if the- ] average temperature reaches 280'F, manually initiate drywell cooling (containment sprays) as described in ONP 525A to b reduce temperature as required. i a Review of Licensee's Submittal + The licensee's calculations for the SB0 room heat-up were reviewed. This plant requires an 8 hour SB0 coping. duration. Therefore, the NUMARC method for calculating dominant area of concern ambient air temperature can not be used. The licensee 17 ~

i l ) .O used a number of alternative methods for calculating room ambient i air temperature for the 8 hour duration. With the exception of the control room, the licensee uf,d a method developed by its-contractor, Devonrue, (15) for longer than four hour duration room l heat up. This method is based on fundamental heat transfer principles and accounts for outside wall temperature' rise of more f tha'n 2.5'K which is the limitation of the NUMARC method. It consists of an' equation for calculating room temperature and { another equation for ensuring that the assumptions in the first ) equation are valid with respect to the heat transfer state within l the wall. We reviewed this method and found it to be applicable for calculating an 8 hour room temperature rise. l The control room heat up calculation is performed by Tenera, a consultant to the licensee, using simultaneous nodal equations for f a transient solution of the heat transfer problems. In this method, materials are modeled as having thermal resistance and ~ j thermal capacitance analogous to the resistance and capacitance in an electrical circuit. Our review of this method is summarized

below under control room evaluation.

1 .The licensee stated that a drywell temperature of 280*F will 3 require contain:nent cooling. The licensee stated that a procedure l 'is,in place which directs the operators to use the diesel driven 1 fire pump to spray water in the drywell using containment spray -nozzles. This requirement to cool the drywell, which is estimated to be required within'the eight hours during an SB0 event, could create a possibility that the suppression pool can be everfilled I beyond its high level limit. In response to this concern, the l licensee provided an analysis wi.ich shows that a maximum of about 500 gallons would be sprayed into the drywell. This quantity is insignificant when compared to the normal suppression pool l' inventory of approximately 696,000 gallons. 4 i 18 L i ,c...

a' s. During Ahe site audit review, the licensee stated that in an SB0 event only a local level indicator will be available to ensure the adequacy of shell side water level in the IC. This may require an entry to the reactor building which does not agree with the licensee's statement that the IC area (reactor building elevation 82' 9") does not require any operator entry during an $80 event. The licensee needs to verify that another IC level indicator is available in the control room, or justify why this level indication will not be needed during an 580 event. o Control Room The licensee provided two calculations for the ambient temperature rise in the control room. Both of these calculations indicated that the control room temperature, would not exceed 120'F during an 8 hour $80 period. These two calculations are reviewed below. The first calculation is an analysis performed by Tenera, a consultant to the licensee, as part of Appendix R control. room heat up calculations. The review of this analysis has resulted in the following concerns, o The assumption of 117 watts per person in the control room during an SB0 is not realistic because other analyses (e.g.,- Millstone 2 control room) and generally accepted references (e.g., ASHRAE Handbook) recommend a value in the raphe of 235 250 watts per person, o The Tenera input of 17.16 Btu /sec for total lighting load is inconsistent and non conservative when compared to earlier calculations by Tenera indicating a value of 62.02 Btu /sec; This non conservative inconsistency requires justification. 19 i

h t o The analysis methodology is dependent upon the various f ~ boundary and control room parameters, such as material resistance (R) and capacitance (C). These parameters ar$ p assigned a numerical value in the Tenera calculation, but there is no basis, calculation, input, or assumptions [ inherent in the selection of these parameters presented in the calculation. This information needs to be provided so that the review can be completed. Also, applicable references used to support the calculations would be needed to review this analysis. o The complex model of parallel trains of resistances, capacitances, heat sources, and heat sinks requires that a more detailed listing of all the input parameters to the computer program be provided. L o Tenera's assumption of boundary temperatures is inconsist'ent with an SB0 scenario and in a non-conservative direction. In the second calculation, the licensee evaluated the ambient temperature rise in both the MP1 and MP2 control rooms using the - CONTEMPT computer code. In this calculation, the licensee evaluated the temperature rise in the MP2 control room (which is similar in design, has a larger heat loads, and smaller heat sink surface area than the MP1 control room) u a bounding analysis for the MP1 control room. Although the results of this " bounding" ^ CONTEMPT calculation are a room temperature of approximately. ll5'F, the assumptions and input are not conservative as discussed below. The B hour SB0 control room heat-up calculation was performed by the licensee using the CONTEMPT computer code which is actually a series of computer codes originally developed to perform containment pressure temperature response calculations due to a 20 n-m. w ..-w .ww.

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-} a i ) LOCA.. Use of the CONTEMPT computer code requires the generation of input and making assump',:ons. Many of these inputs'which can directly affect the results of this calculation were not derived or explained in the licensee's calculation. This' includes: initial room humidity, basis and conservatism of time step size, and basis for concrete, sheet rock, and ceiling material thermal ] conductivity and volumetric heat capacity values, j l A review of the assumptions and input to the CONTEMPT control room model has resulted in the following additional specific concerns: j i o The outside wall boundary temperatures are non-f conservatively low for an SB0 event. These temperatures should be higher and the basis for their values must be justified. c-o The control room initial temperature is non conservative 1'y low, it should have been the maximum allowable temperature. -instead of a typical value of 73'F. I o The values.for thermal conductivity and volumetric specific heat of the concrete, wa11 board, and ceiling tiles which are modeled as heat sinks in the CONTEMPT calculation are non conservatively large. The licensee uses values of 1.04 3 (Btu /hr.ft.'F) and 35.86 (Btu /ft.'F) for concrete thermal conductivity and specific heat respectively. Most literature and containment analyses use values of i 7 3 (Btu /hr.ft.'F) and 28 (btu /ft.*F) for these two parameters. The licensee also uses a thermal conductivity of 4.3 L (Btu /hr.ft.'F) for the sheet rock and ceiling materials which is a relatively high value for materials that are { designed to provide insulation. This value of 4.3 needs to ~ be justified or a lower, more defensible, value should be used. For both materials, the use of lower values of i 1 21 , - - ~, ~ -, - - -..

p 4 thersel conductivity and specific heat will reduce the - energy absorbed by the wall and ceiling heat sinks and increase the SB0 room temperature. o The CONTEMPT code model used by the licensee assumes a heat transfer coefficient of 2 (8tu/hr.ft'.'F) between the room atmosphere and the heat sink surface.. Typical natural convection heat transfer coefficients are usually _about 1 (Btu /hr.fta,p), 5. Containment Isolation Licensee's Submittal The licensee reviewed the plant list of containment isolation valves to verify that valves which must be capable of being closed- ~ or of being coerated (cycled) under SB0 conditions can be positioned (with indication) independent of the prefeared and' blacked-out unit's class 1E power supplies. The licer.see has stated that the 1'o110 wing procedure change will be' implemented to ensure that the containment integrity is provided under SB0 conditions (16): o Procedure ONP 503C will be modified to address the required operator actions to close the following valves <in: order to ensure containment integrity: a. Remote operated DC valves: IC-2, C 3, SD 2A, SD 2B, CV 3, C0-5 and RC-15, b. Local manual operation: LP 2A to LP-20,CS-2A and 1 CS 28. ?2

i; J ,3 Review of Licensee's submittal We performed an independent review of the plant CIVs, provided by the licensee, to identify CIVs requiring either manual or power operated closure capability by excluding those CIVs that conform to the criterion stated in RG 1.155, Section 3.2.7. Our review concurs with the licensee's conclusions that no plant modification I is required to ensure containment integrity. l 6. Reactor Coolant Inventory l Licensee's Submittal The licensee stated that the AAC power source provides power for systems to maintain adequate reactor coolant syttom (RCS) in.entory to ensure.that the core is cooled for the required i coping duration. The licensee analyzed two postulated bounding' scenarios to maximize the extent of core uncovery during an S80. In both cases the licensee assumed a continuous RCS leakage rate l of 75 gpm which is comprised of a 25 gpm per pump for each of the i two reactor recirculation pumps in accordance with NUMARC 87-00 guidelines and a total of 25 gpm for allowed RCS leakage per plant Technical Specifications, Section 3.6,0. In addition, the licensee asramed that the IC is activated on high RCS' pressure at one minutt from the onset of the accident. During the first minute, the licensee estimated the RCS inventory loss through the safety valves. After that, the licensee assumed no RCS inventory loss through safety valves until the automatic depressurization system (ADS)isactuated. In the first scenario, the licensee assumed that the operators would keep the reactor pressure at 900 psia by removing decay heat using the IC. The licensee calculated that with no ADS actuation and RCS make-up it would take three hours for core uncovery to 23

-.~ q. c.' occur. However, if after one hour the ADS is actuated to allow the i low pressure diesel driven fire pump to inject water into the reactor vessel, the resultant RCS shrinkage muld cause a momentary 1% core uncovery just before fire pump core make-up flow CoIWhences. ) In the second scenario, the licensee assumed that the operators x. would use the IC to cool.down the reactor until the reactor vessel water level reaches the top of the active fuel and then actuate l the ADS valves. In this case, the calculations show the combined-l effects of the RCS shrinkage and the inventory loss due to the l actuation of the ADS valves would cause a momentary core uncovery of 33.1/3% before fire pump core make up is initiated. The licensee stated that a momentary 1/3 core uncovery would not result in core damage since this level corresponds to the top of j the jet pumps which is a typical LOCA reflood levels. The [ licensee'also stated that the additional water available outsid'e the core shroud which was not included in this calculation would signif.icantly increase the minimum core level. Finally, the licensee stated that the control rod drive mechanism-I (CRD) pump, wh'ich has a capacity of 77 gpm at high reactor pressure and is available after one hour from the onset of an SB0 l event using AAC source, will provide RCS make up to replace the assumed 75 gpm RCS inventory loss. At low pressure the CRD pump l ~ has sufficient capacity to compensate for any RCS shrinkage and i L-losses through the ADS valves to preclude any core uncovery. L Review of Licensee's Submittal I We reviewed the licensee's calculations by evaluating the licensee's assumptions, input parameters, and the methodology used to arrive at the aforementioned stated results. We found the licensee's calculations for both core uncovery scenarios to be 24 .,.n...- -e..,.., ,n-..,,,.,,,, .. _n,,...-n.

~ 3

j. ;

~ j 1 I acceptable. Furthermore, it was determined that suppression pool cooling would not be required for the two bounding 580 scenarios in the licensee's submittal. 3.4 Proposed Procedure and Training Licensee's submittal The licensee has. indicated that the plant procedures have been reviewed .] .to the guidelines in NUMARC 87-00 for. g; { M 1. station blackout response procedure, ) l. l 2. severe weather procedure, and j i u-3. AC power restoration procedure. f E The licensee has listed-the names of the plant precedures which fall in each of the above categories in the plant SB0 submittal. The. licensee 1 also provided us with an expanded description of the required 580 li procedure revisions and stated that any changes necessary to meet the NUMARC guidelines will be implemented within'one year after NRC notification.is provided.. Review of Licensee's Submittal Our review did not examine the affected SB0 procedures or training. We-view these, procedures as plant specific actionis concerning the required i activities to cope with an SBO. We believe it is the licensee's responsibility to revise and implement these procedures, as needed, to mitigate-an SB0 event. and to assure that those procedures are complete and correct and that the associated training needs-are carried out -j accordingly. 25 i

[r j r 3.5 Proposed Modifications Licensee's submittal ] The licensee has concluded that no hardware modifications for SB0 equipment are necessary to cope with the minimum acceptable 580 duration of eight hours. Review of Licensee's submitta) We found that the present location of 14H bus is vulnerable to the same weather related conditions that could cause loss of offsite power at the site. In addition, the power cable from 14H bus to MP2 emergency bus is partially exposed to the same conditions. Therefore, compliance with. the paragraph B.3 of Appendix B of NUMARC 87 00 requires modifications for MP1 to use the MP2 EDGs as an AAC power source. t 3.6 Quality assurance And Technical Specifications quality Assurance The SB0 equipment list provided by the licensee indicates that all the mechanical equipment is safety related. The equipment is classified as category 1 or fire protection, and is covered by appropriate Appendix B or Appendix R QA requirements. In the SB0 instrument list, the licensee l identified the following as non-QA. i l 1. RPV level: LI-640-29A, L/FR 640-26 and LI-640 29B, and 2. RPV pressure: PI-640-25A, P/FR-640-27 and PI 640-25B The licensee has stated that the non-QA equipment used for SB0 will be covered under an appropriate QA program consistent with the guidance of RG 1.155, Appendix A (17). l' l 26 I

I We agree with the licensee's proposed. action which is consistent with the guidance of RG 1.155 and the requirements of the SB0 rule. Technical Snecificatigi _ i i The category 1 SB0 equipment is already covered by the plant technical specifications. The fire protection and other non-class 1E equipment used to cope with an SB0 should also be covered by appropriate technical ) specifications consistent with the guidance of RG 1.155, Appendices A and B. l l 1 l I l L l 1 27 l l

.I i ts 4.0 CONCLU$10NS I Based on our review of the licensee's submittals and the related supporting documents and discussions during a site audit for the Millstone Point Nuclear Power Station Unit No, I (MP1), we find that the submittal-conforms with the requirements of the SB0 rule and the guidance of RG 1.155 with the following exceptions: I 1. Alternate AC Power Source The 1ccation of the 14H bus, the AAC cross tie bus between MP1 and j MP2, makes it vulnerable-to conditions (i.e. salt spray, flood, ) hurricane, and tornado) that could ca9se a total loss of offsite I power at the site. Even if the bus enclosure (cubicle) could structurally survive the extremely severe wer.ther conditions, it is subject to electrical failures due to water ingress into the, o cubicle during entering and leaving the cubicle for local breaker j operations and from normal water seepage through the walls. In ] L addition, the power cables cross connecting between the units are partially exposed to the same conditions as those of the cubicle. 1 n Also, the structural integrity of both the cubicle and the power j f. cable supports-should be' appropriately evaluated. It is necessary. to provide protection for AAC components by enclosing the 14H Bus in a structure and burying the power cables between the buildings to comply with paragraph B.3 of Appendix B to NUMARC 87-00. i l 2. Class IE Battery capacity I The licensee's calculations of the battery loading during an SB0 is based on the actual required current (ammeter readings) instead of the nameplate ratings. We believe tbt the use of the actual i ammeter readings conforms to the guidance for present capacity calculations provided that: 28 i l l

.~. -- q ) 1. The ammeter readings represent the maximum values taken over i a period of testing and not just from a one time test. '2. The licensee re evaluate the battery capacity if any change to the present plant DC loading occurs, 3. The normal battery backed plant monitoring and electrical-system controls in the control room remain operational during an SB0 event. These are considered to be essential I for successful. coping with and recovery from an SB0 as j documented in NUMARC 87-00 Supplemental Questions / Answers. i 3. Loss of Ventilation l a. Reactor buildina - Elevation B2' 9" ~ During the site audit review, the licensee stated that in an SB0 event a local level indicator will be available to ensure the I adequacy of shell-side water level in the IC. This does not agree with the licensee's statement that the IC area (reactor building elevation 82' g") does not require any operator entry during an ,SB0 event. The licensee needs to provide clarification for the above inconsistency. b.. Control Room ) The licensee provided two calculations for the ambient temperature rise in the control room. Both of these calculations indicated that the control room temperature would not exceed 120'F during an L 8-hour SB0 period. However, there are major problems with both calculations that invalidates the stated results. This problems l need to be resolved. A summary of concerns are stated in Section 3.3, part 4 of the text. y j' LL - - - + - _, - - - - -e ,-w -w---- ,-s e,, ~. - -*w w-ws-e-w~w v-mv-

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u. 3 4. -Quality Assurance and Technical Specifications Quality Assurance Procram The licensee is committed to provide an appropriate QA program for the following non-QA instrumentation consistent with the guidance of RG 1.155, Appendix A: 1 a. RPV 1cvel: L1-640 29A, L/FR 640 26 and LI 640-298, and b. RPV pressure: PI 640 25A, P/FR 640 27 and PI 640-258 .j l Technical Snecifications a To comply with the guidance of RG 1.155, Appendices A and B it is necessary to provide appropriate technical specifications for the fire protection.and other non-class IE equipment used to cope with an SBO. l t F b 30 .. = - _.2.

. ~ _. _ l 6 5.0 References i 1. The Office of Federal Register, " Code of Federal Regulations Title 10 j l Part 50.63," 10 CFR 50.63, January 1, 1989. t \\ . 2. U.S. Nuclear Regulatory Comission, " Evaluation of Station Blackout l Accidents at Nuclear Power Plants Technical Findings Related To Unresolved Safety Issue A 41," NUREG 1032, P. W. Baranowsky, June 19.88. 3. U.S. Nuclear Regulatory Comission, " Collection' and Evaluation of ) Corplete and Partial losses of Offsite Power at Nuclear Power Plants," i tREG/CR 3992, February 1985, i 4. U.S. Nuclear Regulatory Comission, " Reliability of Emergency AC Power ] System at Nuclear Power Plants," NUREG/CR 2989, July 1983. j i

5. -

U.S. Nuclear Regulatory Comission, " Emergency Diesel Generator l Operating Experience, 1981 1983," NUREG/CR 4347, December 1985. 6.- 'U.S. Nuclear Regulatory Comission, " Station Blackout Accident Analysts I l (Part of NRC Task Action Plan A-44)," NUREG/CR-3226, May 1983. 7. U.S. Nuclear. Regulatory Comission Office of Nuclear Regulatory Research, ' Regulatory Guide 1.155 Station Blackout," August 1988. 8. Nuclear Management and Resources Council, Inc., " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blaceit at Light Water Reactors," NUMARC 87-00, November 1987. ] + (. i 9. Nuclear Safety Analysis Center, 'The Reliability of Emergency Diesel Generators at U.S. Nuclear Power Plants," NSAC-108, Wyckoff, H., L September 1986. l L l g

L L h 10. Mroczka, E. J., letter to T. E. Murley of U. S. Nuclear Commission, ' Millstone Nuclear Power Station, Unit Nos.1, 2, and 3 Response to i Station Blackout Rule," Docket Nos. 50-245,50-336,50-423,(B13180), 1 dated April 17, 1989. 11. CALC-1-ENG-89 1 Rev. O, ' Millstone Unit 1 Station Blackout Project j Initiative #1 - Risk Reduction Required Coping Duration Category," by K. E. Murphy, dated March 13, 1989. 1 12. " Millstone Unit 1, 2, and 3, and Connecticut Yankee Plant Emergency j Diesel Generator Reliability data," Provided by M. Marino of NUSCO Operation Department on July 18, 1989. i 1 l 13. U.S. Nuclear Regulatory Commission, " Standard Technical Review Plan for l. the Review of Safety Analysis Reports for Nuclear Power Plants LWR Edition," NUREG-0800, June 1987. L 14. MP1 Drawing No. 25202-26012 Sheet 2 Rev. 6 "P&ID for Main Steam I System," dated January 13, 1989. 15. Mroczka', E. J., letter to T. E. Murley of U. S. Nuclear Regulatory Commission, ' Millstone Nuclear Power Station, Unit Nos.1, 2, and 3 Response to Station Blackout Rule--Additional information," Docket Nos. 50 245, 50-336, 40-423 (B13267), dated May 30, 1989. -16. Murphy, K. E., ' Millstone Unit 1 Station Blackout Procedure Compliance J Review," dated July 19, 1989. J 17. Northeast Utilities Inter-Office Memorandum from R. C. Thomas to G. E. Cornelious, ' Quality Assurance Requirements For Station Blackout (SBO)- All Units," GMB 89-441, dated August 11, 1989, i 1 1 32 l l.-

,t -18. Thadant, A. C., letter with attachment 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.) 19. Thadant, A. C., letter with attachment to W. H. Rasin of NUMARC, " Approval of NUMARC Documents on Station Blackout (TAC-40577)," dated October 7, 1988. 20. Mroczka, E. J., letter to T. E. Murley, "Haddam Neck Plant, Millstone Nuclear Power Station Units Nos. 1, 2, and 3, Response to Station Blackout, Additional information," Docket Nos. 50-213, 50 245, 50-336, 50 423, dated March 30, 1990. 21. Tam,~.P. S., memorandum for, " Daily Highlight - Forthcoming Meeting with ' NUMARC on 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. 22. Rosa, F., Memorandum to Docket Concerning Beaver Valley Units 1 and 2 " Meeting Summary - Meeting of February'22, 1990, on Station Blackout . Issues (TAC 68510/68511)," Docket Nos. 50-334 and 50-412, dated March 6, 1990. 23. Russell, W. T., letter to W. Rasin of NUMARC, " STATION BLACK 0UT," dated June 6, 1990, 33 1

) t t i u J 4 4.0 CONCLU$ IONS Based on our review of the licensee's submittals and the related ] supporting documents and discussions dering a site audit for the Millstone Point Nuclear Power Station Unit No.1 (MP1), we find that the submittal conforms with the requirements of the SB0 rule and the guidance of RG 1.155 with the following exceptions: 1. Alternate AC Power Source i The location of the 14H bus, the AAC cross tie bus between MP1 and MP2, makes it vulnerable to conditions (i.e. salt spray, flood, hurricane, and tornado) that could cause a total loss of offsite power at the site. Even if the bus enclosure (cubicle) could j structurally survive the extremely severe weather conditions, it is subject to electrical failures due to water ingress into the, cubicle during entering and leaving the cubicle for local breaker operations and'from normal water seepage through the walls. In addition, the power cables cross connecting between the units are partially exposed to the same conditions as those of the cubicle. L Also, the structural integrity of both the cubicle and the power cable supports should be appropriately evaluated. It is necessary to provide protection for AAC components by enclosing the 14H Bus i in a structure and burying the power cables between the buildings to comply with paragraph B.3 of Appendix B to NURARC 87 00,- E 2. Class IE Battery Capacity u 1, The licensee's celculations cf the battery loading during an SB0 is based on the actual required current (ammeter readings) instead- [ of the nameplate ratings. We believe that the use of the actual 4 ammeter readings conforms to the guidance for present capacity calculations provided that: 28 .x, ~

'l<' l. 'The ammeter readings represent the maximum values taken over a period of testing and not just from a one time test. 2. The licensee re evaluate the battery capacity if any change to the present plant DC loading occurs. j s D-3. The normal battery-backed plant monitoring and electrical i system controls in the control room remtin operational during an SB0 event. These are considered to be essential ) for. successful coping with and recovery from an SB0 as documented in NUMARC 87-00 Supplemental Questions / Answers. L 3. Loss of Ventilation a. Reactor buildina - Elevation 82' 9" l 1 ~ During the site audit review, the licensee stated that in an SB0 l event a local level indicator will be available to ensure the adequacy of shell side water level in the IC. This does not agree withthelicensee'sstatementthattheICarea(reactorbuilding elevation 82' g") does not require any operator entry during an SB0 event. The licensee needs to provide clarification for the-aboyc inconsistency. i f f b.- Control Room 1 The licensee provided two calculations for the ambient temperature rise in the control room. Both of these calculations indicated g' .that the control room temperature would not exceed 120'F during an-l 8 hour SB0 period. However, there are major problems with both calculations'that invalidates the stated results. This problems need to be resolved. A summary of concerns are stated in Section 3.3, part 4 of the text. l 29 a ~ .k----.

.,. =,, gf 4. Quality Assurance and Technical specificaticas Quality Assurance Proaram I I The licensee is committed to provide an appropriate QA program for the following non-QA instrumentation consistent with the guidance of RG 1.155, Appendix A: a. RPV ' level-L1 640-29A, L/FR 640 26 and LI 640 29B, and b. RPV pressure: PI-640 25A, P/FR 640-27 and PI-640-25B w Iachnical Soecifications To comply with the guidance of RG 1.155, Appendices A and B it is necessary to provide appropriate technical specifications for the-fire protection and other non-class IE equipment used to cope w'ith l an SBO. + e s i l 30 l l [ '. 1 .... ~ _ _.. _..... _ _... -...... _, _ _, -

p" .= [;. ( 5.0 References I 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 Comission, " Evaluation of Station Blackout Accidents at Nuclear Power Plants - Technical Findings Related To l Unresolved Safety Issue A-44," NUREG 1032, P. W. Baranowsky, June 1988. 3. U.S. Nuclear Regulatory Comission, " Collection and Evaluation of Complete and Partial losses of Offsite Power at Nuclear Power Plants," ~ NUREG/CR 3992, February 1985. -4. U.S. Nuclear Regulatory Comission, " Reliability of Emergency AC Power System at Nuclear Power Plants," NUREG/CR-2989, July 1983. 5. U.S. Nuclear Regulatory Comission, " Emergency Diesel Generator Operating Experience, 1981-1983," NUREG/CR-4347, December 1985. t 6. U.S. Nuclear Regulatory Comission, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," NUREG/CR-3226, May 1983. 7. U.S. Nuclear Regulatory Comission Office of Nuclear Regulatory Research, " Regulatory Guide 1.155 Station Blackout," August 1988. 8. Nuclear Management and Resources Council, Inc., " Guidelines and 1 Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," NUMARC 87-00, November 1987. h l-9. Nuclear Safety Analysis Center, "The Reliability of Emergency Diesel Generators at U.S. Nuclear Power Plants," NSAC-108,'Wyckoff, H., September 1986. I ~ t 31 l\\I ~

2 . o, .g 10. - Mroczka E. J., letter to T. E. Murley of U.' S. Nuclear Commission, " Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3 Response to Station Blackout Rule," Docket Nos. 50 245, 50-336, 50 423, (813180), dated April 17, 1989.

11. -

CALC 1-ENG 89-1 Rev. O, " Millstone Unit 1 Station Blackout Project Initiative #1 - Risk Reduction Required Coping Duration Category," by 4 K. E. Murphy, dated March 13, 1989. 12. " Millstone Unit 1, 2, and 3, and Connecticut Yankee Plant Emergency Diesel Generator Reliability data," Provided by M. Marino of NUSCO Operation Department on July 18, 1989. 13. U.S. Nuclear Regulatory Commission, " Standard Technical Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants LWR Edition," NUREG 0800, June 1987. 14. MPl Drawing No. 25202-26012 Sheet 2 Rev. 6 "P&lD for Main Steam System," dated January 13, 1989. 15. Mroczka, E. J., letter to T. E. Murley of U,. S. Nuclear Regulatory Commission, " Millstone ~ Nuclear Power Station. Unit Nos. 1, 2,'and 3 Response to Station Blackout Rule--Additional information," Docket Nos. 50-245,50-336,40-423,(813267), e.ied May 30, 1989. 16. Murphy, K. E., " Millstone Unit 1 Station Blackout Procedure Compliance Review," dated July 19, 1989. 17. Northeast Utilities Inter-Office Memorandum from R. C. Thomas to G. E. Cornelious, " Quality Assurance Requirements For Station Blackout (SBO)- All Units," GMB 89 441, dated August li, 1989. 32

' '). Thadani, A. C., letter with attachment to A. Marion of NUMARC, " Publicly We+ iced Mei ng December 27, 1989," dated January 3, 1990 (Confirming

  • NUMARC 87-00 Supplemental Questions / Answers," December 27,1989.)

19. Thadani, A. C., letter with attachment to W. H. Rasin of NUMARC, " Approval of NUMARC Documents on Station Blackout (TAC-40577)," dated October 7, 1988. ? 20. Mroczka, E. J., letter to T. E. Murley, "Haddam Neck Plant, Millstone 6 Nuclear Power Station Units Nos. 1, 2, and 3. Response to Station Blackout, Additional information," Docket Nos. 50 213, 50 245, 50-336, 50 423, dated March 30, 1990. 21. Tam, P. S., memorandum for, " Daily Highlight - Forthcoming Meeting with NUMARC on 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. 22. Rosa, F., Memorandum to Docket Concerning Beaver Valley Units 1 and 2, " Meeting Summary - Meeting of February 22, 1990, on Station Blackout issues (TAC 68510/68511), Docket Nos. 50 334 and 50-412, dated March 6, 1990. 23. Russell, W -T., letter to W. Rasin of NUMARC, " STATION BLACKOUT," dated June 6, 1990. 33 w - -.--- :}}

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