ML20063P933
| ML20063P933 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/18/1990 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20063P934 | List: |
| References | |
| CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-89-1151, TAC-68568, NUDOCS 9008030198 | |
| Download: ML20063P933 (29) | |
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ATTACHMENT f l
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s-I TECNNICAL IVALUATION REPORT h
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NILLSTONE NUCLEAR POWER STATION UNIT N0. 3 i
STATION BLACK 0UT EVALUATION
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SAIC l
Science Applicaticns IntemationalCorparation l
An Employee Owned Company t
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1 ATTACHMENT 1 i
$A!C-89/1151 j
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TECHNICAL EVALUATION REPORT i
MILLSTONE NUCLEAR POWER STATION UNIT NO. 3 I
i STATION BLACK 0UT EVALUATION i
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TAC N0. 68568 l
5AIC Final July 18,1990 Prepared for:
U. S. Nuclear Regulatory Comission Washington, D. C. 20555 Contract NRC 03 87-029 Task Order No. 38 Pcet Offee Boa 13D,1710 Goodrktge Drin McLeon. Yvgir4 22102. (7@ 8?l G0
1 TABLE OF CONTDfTS Section falt
1.0 BACKGROUND
1 2.0 REVIEW PROCESS........................................
3 3.0 EVALUATION............................................
6 3.1 Proposed Station Blackout Duration...............
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3.2 Alternate AC Power Source........................
8 3.3 Station Blackout Coping Capability...............
9 3.4 Proposed Procedures and Training.................
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l 3.5 Proposed Modifications...........................
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L 3.6 Quality Assurance and Technical Specifications...
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4.0 CONCLUSION
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5.0 REFERENCES
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u TECHNICAL EVALUATION REPORT MILLSTONE NUCLEAR POWER PLANT, UNIT No. 3
$TATION BLACK 0UT EVALUATION 1.0 SACKGROUNO 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 i
Alternating Current Power" (1). The objectivt of this requirement is to assure 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 reouired duration. This requirement is based on information developed under the commission study of Unresolved Safety Issue A 44, " 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 j
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 NUMARC 87 00 and concluded that the NUMARC document provides an acceptable guidance for addressing the 10 CFR 50.63 requirements. The application of this methoi results in selecting a minimum acceptable SB0 duration capability i
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.
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In order to achieve a consistent systematic response from licensees to the SB0 rule and to expedite the staff review process, NUMARC developed two generic response documents. These documents were reviewed and endorsed by the-e NRC staff (19) for the purposes of plant specific submittals. The documents j
are titled:
1.
" Generic Response to Station Blackout Rule for Plants Using Alternate AC Power," and i
i 2.
" Generic Response to Station Blackout Rule for Plants Using AC Independent Station Blackout Response Power."
1 A plant-specific submittal, using one of the above generic femats, J
provides only a summary of results of the analysis of the plant's station blackout coping capability.
Licensees are expected to ensure that the
'l baseline assumptions used in NUMARC 87-00 are applicable to their' plants and to verify the accuracy of the stated results.
Compliance with the SB0 rule' requirements is verified by review and evaluation of the licensee's submittal-and audit review of the supporting documents as necessary.
Follow up NRC inspections assure that the licensee has implemented the nectssary changes as required to meet the SB0 rule.
E in 1989, a joint NRC/SAIC team headed by an NRC staff member performed audit reviews of the methodology and documentation that support the licensees' submittals for several plants.
These audits revealed several deficiencies which were not apparent from the review of the licensees' submittals using the agreed upon generic response format. These deficiencies raised a generic l
question regarding the degree of licensees' conformance to the requirements of L
the SB0 rule. To res *t: inis question, on January 4, 1990, NUMARC issued additional guidance as NUMARC 87-00 Supplemental Questions / Answers (20) addressing the NRC's concerns regarding the deficiencies.
NUMARC requested
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that the licensees send their supplemental responses to the NRC addressing these concerns by March 30, 1990.
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2.0 REVIEW PROCE$$
The review of the. licensee's submittal is focused on the following areas consistent with the positions of RG 1.155:
A.
Minimum acceptable SB0 daration (Section 3.1),
B.
SB0 coping capability (Section 3.2),
C.
Procedures and training for $80 (Section 3.4),
D.
Proposed modifications (Section 3.3), and E.
Quality assurance and technica~1 specifications for SB0 equipment (Section 3.5).
l For the determination of the proposed minimum acceptable 5B0 duration',
the following factors in the licensee's submittal are reviewed:
a)offsite power design characteristics, b) emergency AC power system configt ition, c) determination of the emergency diesel generator (EDG) reliability consistent with NSAC-108 criteria (9), and d) determination of the accepted EDG target reliability. Once these factors are known, Table 3 8 of NUMARC 87-00 or Table 2 of RG 1.155 provides a matrix for determining the required coping duration.
for the 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 at,ove. The review process follows the guidelines given in RG 1.155, Section 3.2, to assure:
q a.
availability of sufficient condensate inventory for decay heat-
- removal, 3
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adequacy of the class 1E battery capacity to support safe
- shutdown, c.
availability of adequate compressed air for air operated valves necessary for safe shutdown, 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 f.
ability of the plant to maintain adequate reactor coolant system inventory to ensure core cooling for the required coping duration.
The licensee's submittal is reviewed to verify that required procedures
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(i.e., revised existing and new) for coping with SB0 are identified and that appropriate _ operator training will be provided.
The licensee's submittal for any proposed modifications to emergency AC sources, battery capacity, condensate capacity, compressed air capacity, appropriate containment integrity and primary coolant make up capability is reviewed.
Technical sp:cifications and quality assurance set forth by the licensee to ensure high reliability of the equipmert, specifically added or assigned to meet the requirements of the SB0 rule, are assessed for their adequacy.
The licensee's proposed use of an alternate AC power source is reviewed to determine whether it meets the criteria and guidelines of Section 3.3.5 of RG 1.155 and Appendix B of NUMARC 87 00.
A normal SB0 review is limited to the review of the licensee submittal; it does not include a concurrent site audit review uf the supporting documentation.
Such an audit may be warranted as an additional confirmatory 4
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action..This determination would be made and the audit would be scheduled and performed by the NRC staff at some later date.
logo However, a limited number of concurrent site audit reviews were performed in order to obtain a benchmark for licensee conformance with the q
documentation requirements of the SB0 rule. Millstone Nuclear Power Station i
Unit No. 3 was one of the plants selected by the NRC staff for a concurrent audit review of the SB0 supporting documentation. This audit was performed by the 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 (14), and March 30, 1990 (18), and the site audit review.
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i 3.0 EVALUATION 3.1 Proposed Station Blackout Duration Licensee's Submittal The licensee, Northeast Utilitie> ',NU), calculated (10,14 and 18) a minimum acceptable SB0 duration of eight hours for the Millstone Nuclear Power Station Unit No. 3 (MP3). The licensee stated that no modifications are necessary to attain this proposed coping duration.
The plant factors used to estimate the proposed $80 duration are:
1.
Offsite Power Design Characteristics The plant AC power design characteristic group is *P3" based on:
a.
Independence of the plant offsite power system characteristics of "il/2."
b.
Expected frequency of grid-related LOOPS of less than one per 20 years, c.
Estimated frequency of LOOPS due to extremely severe weather of 1.20E-02 per year (10) which places the plant in ESW group "5,"
and d.
Estimated frequency of LOOPS due to severe weather of 1.47E 01 per year which places the plant in SW grocp "5."
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EmergencyAC(EAC)PowerConfigurationtroup The EAC power configuration of the plant is "C."
MP3 is equipped with two emergency diesel generators, which are normally available to the unit safe shutdown equipment.
The licensee stated that one EAC power supply is necessary to operate safe shutdown equipment following a LOOP.
3.
Target Emergency Diesel Generator (EDG) Reliability The licensee has selected a target EDG reliability of 0.975 based on having a nuclear unit average reliability for the last 100 demands greater than 0.95 consistent with NUMARC 87 00 selection criterion.
Review of Licensee's submittal
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r Our review concurs with the evaluation of the proposed SB0 coping duration.
Factors which affect the estimation of the $80 cop,ing 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.
The licensee's estimation of the frequency of LOOPS due to ESW and SW was based on the data provided in NUMARC 87-00.
The independence of offsite power system grouping at MP3 is "I 1/2." Ali the offsite tower sources at MP3 are connected to the unit's safe shutdown buses through one switchyard. The normal soure of AC power to these buses are from the unit main generator 10.d there is an automatic transfer (generator H
breaker) to one preferred power source.
If this preferred power source fails, there is another automatic transfer (fast transfer) to the remaining preferred power sources or to alternate offsite power source.
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The assignment of the EDG target reliability is supported by the demonstrated 100% start and load-run EDG reliability over the last 100 demands (11). No detailed review of the EDG reliability was performed during the site audit. The licensee stated that the present EDG 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 added, however, that a formal reliability program will be established in accordance with the above guidance, and the targeted EDG reliability will be maintained (18).
3.2 AlternateAC(AAC)Powersource Licensee's submittal The licensee's submittal dated March 30, 1990 (18), stated that the AAC power source for MP3 will be a dedicated air cooled diesel generator, of sufficient capacity, to provide power for the unit's $80 loads.
The' licensee is in the process of developing the details of this design.
The licensee stated that the AAC power source will meet the criteria i
specified in Appendix B to NUMARC 87-00 and will be available within one hour from the onset of an SB0 event.
Review of Licensee's $ubmittal The installation of an independent air cooled diesel generator (DG) as an AAC power source is different from that which was evaluated during the site audit review in July of 1989.
Since the details of the AAC power source were not provided, no review can be performed.
During the site audit review, the licensee's identified SB0 loads for MP3 were 663 kW. This estimah was based on the actual required loads instead of the nameplate ratings. The licensee needs to consider a DG of sufficient capacity to power the comittM loads (using nameplata ratings) as documented in the Unit's SB0 submittals (10, 14 and 18).
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4 3.3 Station Blackout coping Capability i
The plant coping capability with an SB0 for the required duration of eitM hours is assessed based on the following results:
j 1.
Condensate Inventory for Decay Heat Removal j
Licensee's submittal The licensee's submittal stated that a total of 166,000 gallons of water are required for decay heat removal and reactor cooldown for the proposed SB0 duration of eight hours. The minimum permissible condensate level for the demineralized water storage tank per L
technical specifications provides 334,000 gallons of water, which I
exceeds the requ' ired quantity for coping with an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5B0.
The I
licensee stated that no plant modifications or procedural changes are necessary to use this water source.
Review of Licensee's Submittal j
The needed condensate inventory for decay heat removal was calculated by the licensee using an in house computer code, THIST, which estimates the integrated decay heat release during an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> SB0 duration. The licensee stated that this computer code has been qualified for safety related cair.ulations.
For the purpose of this review, w used data from Branch Technical Position ASB 9 2, 'Resideal Decay Heat Release Rate for Light Water Reactors" (12) to estimate an eight hour integrated decay heat similar to the licensee's calculated value.
The reactor cooldown to approximately 364'F was calculated by the licensee with an analysis of the stored energy of all the water, fuel, and metal in the primary coolant system using data frcm the plant containment analysis. The stored energy and cooldown calculations 9
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were reviewed and found to be correct. We agree with the licensee r
that the plant has an. adequate condensate inventory to cope and I
recover from an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> $80 event.
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J 2.
Class IE Battery Capacity Licensee's Submittal i
1 The licensee stated that a calculation has been performed to
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verify that the class 1E station batteries have sufficient j
capacity to meet the required $80 loads for one hour. After the initial one hour, when the AAC power source is available, the chargers for these batteries will be powered from the AAC_ power j
source (13).
Review of Licensee's 3ubmittal During the site audit review, the licensee provided an analysis of capacity calculation for battery "1" and demonstrated that it has a capability to supply the SB0 loads for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The l' ensee l
stated that since the loads are almost equally distributed between the class IE battery "1" and "2." this analysis is also applicable to battery "2."
l The licensee's battery capacity calculations is based on actual ammeter readings on inverters and battery chargers during a loss of power test.
The licensee's test data indicates that the battery charger load is 50 amperes before a reactor trip. The licensee added this current (the charger ammeter reading) to the l
inverter current readfi1 after the reactor trip for an estimate of i
l the continuous loading of the battery.
The test data, which was l
taken in March of 1986, shows a zero current on the inverter prior to the reactor trip. After the reactor trip, the current jumped L
to 65 amperes, and when the battery charger was manually isolated, I
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the inverter current reached 70 amperes.
In another ammeter reading, in support of $80 electrical heat loads on the same battery charger and inverter during normal plant power operation on March 10, 1989 (21), the current on the charger and inverter were 40 and 33 amperes, respectively. These two readings, (readings from March 1986 and 1989), show an inconsistency in the inverter current during normal power operation.
If we were to use the-1989 readings and assume that this inverter would need to carry an additional 70 amperes after the LOOP, then the total current on battery "1" should be 143 amperes instead of the 120 amperes which was used in the battery capacity calculations.
If the licensee were to use the latter (143 amperes), we have estimated that the battery may last for eight hours without being charged. However, since the licensee has stated that the battery charger will be powered after one hour when the AAC power is established, we consider this difference in the current to be
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unimportant.
We believe that the use of the actual ammeter readings conforms to the guidance for present capacity calculations provided,that:
1.
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-evaluata 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 $80 as documented in NUMARC 87 00 Supplemental Questions / Answers.
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3.
Compressed Air Licensee's submittal The licensee stated that no air operated valves are relied upon to j
cope with an SB0 for eight hours. The air operated valves I
required for the operation of decay heat removal, i.e. the atmospheric dump valves (ADVs) are operated manually. The 1
auxiliary feedwater (AFW) flow control valves and the steam admtssion valves are powered from class 1E station batteries. The licensee added that shcrt term decay heat removal is accomplished via AFW turbine driven feed pump (TDFP) and the steam generator safety valves.
Long-term decay heat release is accomplished via AFW TOFP and the manual operation of ADVs.
1 Review of Licensee's Submittal J
The licensee's proposed method of decay heat removal is beyond the normal operational envelope.
It requires close coordination of at least two to three operators:
one in the control room to monitor j
both the steam generator levels and the reactor coolant hot and I
cold leg temperatures, and the second and third operators in the j
main steam safety and atmospheric dump valve enclosure building to modulate one or more F
In order to ensure proper operation, this decay heat removal o,.
tion scenario needs to be simulated and operators need to be tra,ied appropriately.
4.
Effects of Loss of Ventilation l
Licensee's submittal The calculated post-SB0 steady state ambient air temperature for the plant areas containing SB0 equipment are as follows (14):
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Area Temnerature f'F) final Initia1 Control Room 107 75 1&C Rack Room 116 75 Steam Driven AFW Pump Room 140 104 Main Steam Valve Butiding 152 104 East DC Switchgear Room 114 85 West DC Switchgear Room g5 85 Charging Pump Cubicles 219 104 The licensee stated that those areas having a final ambient
%;erature greater than 120'F are considered to be a dominant artra of concern (DAC).
Therefore, this eliminates all rooms but the steam driven auxiliary feedwater (AFW) pump room, main steam valve build'ng and charging pump cubicles. The equipment in the steam driven AFW Pump Room has already been qualified for operation during postulated high energy line breaks (HELB) which result in a higher temperature and a harsher environment than that of an SB0 event.
To provide reasonable assurance of equipment operability. the licensee has committed (14) to revising plant SB0 procedures to restart the following HVAC equipment once the AAC power source is established:
1.
Main Steam Valve Building exhaust fan Nos. 3HVV*FNIC or D.
2.
Charging Pump Cubicle Supply and Exhaust Fan Nos.
3HVR*FN13 A or B and 3HVR*FN14 A or B.
3.
Battery Room Exhaust Fan Nos. 3HVC*FN9A, C, and E or B and D.
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e In addition, at one hour, all non essential electrical equipment in the switchgear and control room areas will be shut off.
Review of 1.icensee's Submittal The licensee's SB0 submittal room heat up calculations were reviewed. This plant requires an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> SB0 coping duration.
Therefore, the NUMARC method for calculating dominant area of concern ambient air temperature can not be used..The licensee used an alternative method for calculating room ambient air temperature for the 8-hour duration. The licensee used a method developed by its contractor, Devonrue, (15) for longer than four hour duration room heat-up. This method is based on fundamental heat transfer principles and accounts for outside wall temperature rise of more than 2.5'K which is the limitation of the NUMARC method.
It consists of an equation for calculating room
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temperature and another equation for ensuring that the assumptions in the first equation are valid with respect to the heat transfer state within the wall.
We reviewed ".:.ls method and found it to be applicable for calculating an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> room temperature rise.
The following summarizes the major areas for which we either have concerns or do not have any information to review:
o Evaluation of Adjacent Room Heat-u,1 The licensee correctly used an appropriate 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> method to calculate the room heat up following an SB0 event lasting longer than four hours (15).
This method inherently includes the assumption that the room's enclosure wall outside temperature.
remains constant during the SB0 time period of interest.
It also requires an examination of all the rooms. adjacent to the room of interest to determine if these rooms would be mbject to an SB0 heat-up.
Any heat-up in a room adjacent to the room of interest would result in an increase in the corresponding common wall 14
surface temperature and therefore invalidate the use of this l,.
equation.
l An evaluation of drawings provided by the licensee indicates that I
the' main steam valve room and the turbine driven AFW pump room each have a wall which is part of the containment building. The equipment in the AFW pump room has been previously qualified to operate in a harsh environment due to an HELB. According to the~
licensee, the calculated AFW room temperature is well below HELB temperature profiles.
Therefore, we believe that the heat transfer from the containment would not significantly increase the room temperature which could result in equipment operability l
problems.
However, the licensee needs to re evaluate the room heat-up calculations for the main steam valve room to account for the containment wall temperature rise.
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o Control Room
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L The licensee calculated the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> SB0 control room heat up using a
L the applicable S hour method.
However, the licensee did not consider the common walls between the control room and the l
adjacent instrument and control (!&C) Rack Room.
The results of 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> room heat-uo calculations for both the Control Room and the Rack Room show that the Control Room reaches 107'F while the 1&C l
Rack Room reaches ll6'F.
Therefore, the I&C Rack Room actually becomes a heat source to the Control Room and should have been accounted for.
In addition, the stated 9.4 kW heat load in the control room does not include any heat generation from personnel 4
in the control room. The personnel eat load will significantly increase the control room heat load and raise the calculated control room temperature.
If the same personnel heat generation rate as used in the Millstone 2 analysis is included, the control room temperature would rise to 111'F.
Our evaluation of the control room heat-up with the additional heat load of human 15
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occupancy and the potential effect of the slightly hotter adjacent Rack Roon resulted in the control room remaining below 120*F.
However, the licensee needs to provide assurance that the control room equipment will remain operational at above normal operating temperature by opening the cabinet doors within 30 minutes of the
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onset of an $B0 event to minimize the hot spots. This action is needed to conform with the guidance provided in NUMARC 87 00, I
Supplemental Questions / Answers.
j The licensee's submittal dated March 30, 1990 (18), stated that it is planning to install acoustic ceiling in the MP3 control room.
Since this installation invalidates the open ceiling assumption in the control room heat up analysis, the licensee needs to revise the control room heat up calculations prior to the completion of this modification.
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5.
Containment Isolation Licensee's submittal The licensee reviewed the plant containment isolation valves (CIVs) to verify that valves which must be capable of being closed or be operated (cycled) under SB0 conditions can be positioned (with indication) independent of the preferred and blacked out unit's class 1E power supplies. The licensee has stated that no plant modification and/or associated procedure changes are necessary to ensure that appropriate containment integrity is provided under SB0 conditions.
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 meet one 1
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a or more of the criterion. stated in RG 1.155, Section 3.2.7.
0ur i
L review concurs with the licensee's conclusions that no plant
- modification i. required to ensure containment integrity.
6.
Reactor coolant Inventory Licensee's submittal The licensee stated that the AAC power source provides power for make up systems to maintain adequate reactor coolant system (RCS) inventory to ensure that the core is cooled for the required coping duration. The licensee analyzed two postulated bounding scenarios to maximize the extent of core uncovery during an 580.
In both cases the licensee assumed a continuous RCS leakage. rate of 112 gpm which is comprised of,a 25 gpm per pump for each of the.
four reactor coolant pumps in accordance with NUMARC 87 00 -
guidelines and a total of 12 gpm for allowed RCS leakage per p1~ ant I
technical specifications, Section 3.4.6.2.
In addition, the licensee stated that the operators would try to cooldown the RCS as directed by emergency operating procedure (EOP) ECA 0.0 to reduce the RCS leakage, and maintain subcooled margin and single phase natural circulation cooling. The cooldown for both cases is assumed to be to shutdown cooling entry conditions, which is an RCS pressure of approximately 220 psia and an RCS average temperature of 356'F.
l In the first scenario the licensee calculated that with no RCS make up it would take six hours and thirty minutes for the vessel level to reach to the top of the active fuel.
In the second i.
scenario, the licensee assumed that one charging pump with a capacity of 100 gpm'will be available one hour after the onset of an SB0 when the AAC power source is established.
In this case the licensee has estimated that it would take almost 71 hours8.217593e-4 days <br />0.0197 hours <br />1.173942e-4 weeks <br />2.70155e-5 months <br /> for the vessel level to reach the top of active fuel.
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Review of Licensee's' Submittal We reviewed the licensee's calculations by evaluating the licensee's assumptions, inputs, and the methodology used to arrive at the stated results. We found the licensee's calculations for both core uncovery scenarios to be appropriate. We agree with the licensee that the plant is capable of maintaining the core covered during an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> !% event once the AAC power source to the plant is restored at one hour and RCS make up is established with one l
charging pump.
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l 3.4 Proposed Procedure and Training Licensee's Submittal
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The licensee has indicated that the plant procedures have been reviewed to the guidelines in NUMARC 87-00, Section'4.0, for:
L 1.
Station blackout response, 2.
Severe weather, and 3.
AC power restoration.
The licusee listed the names of the procedures that fall in each of the above three categories in the plant SB0 submittal.
The licensee stated that any changes necessary to meet the NUMARC guidelines will be i
implemented within one year after the NRC notification is.provided.
Review of Licensee's Submittal Our review did not exam;ne the affected procedures or training.
These procedures are plant specific actions concerning the required activities to cope with an SBO.
It is the licensee's responsibility to revise and implement these procedures, as needed, to mitigate an SB0 event and to 18
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assure that these procedures are complete and correct, and that the associated training needs are carried out accordingly.
3.5 Proposed Modifications Licensee's Submittal J
l lhe licensee has proposed to install an independent air-cot, led diesel l
generator to serve as an AAC power source for MP3.
The licensee is currently developing the details of this modification. The licensee did not provide any other information for us to review.
t Review of Licensee's Submittal We are unable to perform a review of the proposed modification since no information is provided.
3.6 Quality Assurance And Technical Specifications Ouality Assurance The SB0 equipment list p ovided by the licensee indicates that all the i
mechanical equipment is category 1, therefore, it is controlled by an appropriate QA program.
The SB0 instrumentation list, on the other hand, identifies 10instrumentwhicharenon-classlE(16),therefore requiring an appropriate QA program. The licensee has stated (17) 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.
We believe that the licensee's proposed action is consistent with the guidance of RG 1.155 and the requirements of the SB0 rule.
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.,a Technical' Snecifications The category 1 SB0 equipment is already~ covered by the plant technical specifications. The non class 1E equipment used to ce.oe with an CBO, should also be covered by an appropriate technical sporifications-consistent.with the guidance of RG 1,155, Appendices A and B.
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,7 CONCLU$!0NS.-
4.0 Based on our review of the licensee's submittals (10 14, and'18) and
-.the relateu supporting documents and discussions during a site' audit for the Millstone Nuclear Power Station-Unit N0. 3 (MP3), 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
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The li:ensee proposes to use an independent air-cooled diesel
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gen;;rator as-an AAC power source.
Sirce no details on this I
modification are provided, we consider this issue an open item-i pending the licensee's AAC evaluation submittal for the staff's i
- review, i
2, Class IE Battery capacity The licensee's calculations of'the battery loading du,ing an SB0 u '.ased the actual required current (ammeter readings),instead of
- nameplate rating. We believe that the use of the actual ammeter
. readings conforms to the guidance for present capacity calculations provided that:
1.
The ammeter readings represent the maximum values taken over a period of testing,-not from a single test.
2.
The licensee re-evaluate the battr-'s capacity if any change i
in the present plant DC loading occurs, i
3.
The normal battery-backed plant monitoring and electrical system controls in the control rooni remain operational during an SB0 event.
These are cons.dered to be essential 21
E L.
for successful. coping'with; and recovery from an SB0 as documented in NUMARC 87-00 Supplemental-Questions / Answers.
3.
Compressed Air ~
~
The licensee's proposed method of decay heat removal. is-beyond the normal-operational envelope.. -It requires a close coordination of ~
at least two to three operators: one in the control' room to monitor both the steam generator levels and the reactor coolant -
hot and cold leg temperatures, and the second and third operators in the. main steam safety and atmospheric dump valves enclosure building to modulate one or more ADVs. Therefore, the licensee-needs to simulate this scenario and train the operators accordingly.
4.
Loss of Ventilation a.
Evaluation of Adiacent Room Heat-un An evaluation of drawings provided by the licensee, that, show all SB0 rooms and rooms adjacent to these rooms, indicates that the main steam valve room and.the turbine driven AFW pump room each have a wall which is the containment building. The licensee stated that the AFW pump room has been previously qualified to operate in a harsh environment due to an'HELB, therefor 1, the equipment-operability in this should be assured.
T' aeat up. calculations for the main steam valve room needs to be re-evaluated to account for the containment wall temperature rise and the equipment operability re assessed at the new temperature.
b.
Control Room The licensee's calculation of ambient temperature rise in this room did not consider the effect of the common walls between the 22
4,.
control room and the I&C Rack Room. This invalidates the main.
assumption used in the control' room heat up calculations.
In addition, the control-room heat load does not' include any heat
. generation from personnel.. This heat load will be a significant fraction of the estimated load from the panels and lighting. 'The-addition of these heat sources would result'in a higher temperature rise than that 6stimated by the licensee.
- Further, the new licensee's submittal dsted March 30 1990s stated that-installation of acoustic ceiling is planned for the control room and the licensee would revise the temperature rise calculations once the modification is complete. The licensee needs to revise the control room heat-up calculations prior to the completion of the proposed' modification by considering both the above concerns and the acoustic ceiling to ensure that the estimated room i
temperature would not cause any equipment operability problems consistent with the guidance provided in NUMARC 87-00 Supplemental Questions / Answers regarding opening control room cabinet doors.E 5.
Quality. Assurance and Technical specifications Ouality Assurance orocram To comply with the guidance of RG 1.155, Appendix A, it is necessary to provide an~ appropriate QA program for the Non-QA instrumentation-used for S80.
Technical Soecifications.
To comply.with the guidance of RG 1.155, Appendix B it is necessary to provide appropriate Technical Specifications for the fire protection and other non-class 1E equipment used to cope with an SBO.
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5.0 ~ Reforences 1, s I:
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. 1.
- The Office of'Feder% Register, " Code of Federal Regulations Titie 10 Part 50.63,* 10 CTR 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.
I 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 Commission, " Emergency Diesel Genera.
~
p Operating Experience, 1981-1983," NUREG/CR 4347, December 19h.
6.
U.S. Nuclear Regulatory Commission, " Station Blackout Accident' Analyse:
(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.
o 8.
Nuclear Management end Resources Council, Inc., " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," NUMARC 87-00, November 1987.
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9.
Nuclear Safety Analysis Center, "The Reliability of Emergency Diesel Generators at U.S. Nuclear Power Plants," NSAC-108, Wyckoff, H.,
September 1986.
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10.-
Mroczka, E.- J., letter to T. E. Murley of U.S. Nuclear Regulatory =
4
' Comission, " Millstone Nuclear Power Station, Unit Nos. = 1, 2. and 3 Response _to Station Blackout-Rule," Docket Nes. 50 245, 50 336, 50-o 423,(813180), dated April 17, 1989.
~ 1.
" Millstone Unit 1, 2, and 3 and Connecticut Yankee Emergency Diesel 1
Generator Reliability data," provided by M. Marino from Nuclear Operation Department on July 18, 1989.
" '12.
U.S. Nuclear Regulatory Comission, " Standard Technical Review Plan tor
'j v
the Review of Safety Analysis Reports for Nuclear Power Plants LWR Edition," NUREG-0800, June 1987.
13.
Northeast Utilities Inter-0ffice Memoran>
+tachment, from i
S. Striker to M. Whitelaw, "Hillstone Unh 4 Sutwo Blaenout Electrical
. Distribution Equipment," GEE Memo No. 89 301, dated August 11, 198}
14.
Mroczka, E.
J., letter to T.
-iurley of U.S. Nuclear Regulatory Commission, " Millstone Nuclear rower Station, Unit Nos. 1, 2, and 3 Response to Station Blackcut Rule -- Additional Information," Docket Nos. 50 245,'50 336 and 50 423, dated May 30, 1989, d
15.
Pareto, V. E., and S. Maloney letter to M. Childers of Northeast-Utilities, " Thermal Stratification-on the Walls of Domioar.t Areas of Concerns,." DMSBO-88 12 02/VEP, dated December 19, 1988.
16.-
Northeast titilities Inter-Office Memorandum..with Attachment,~ from 1
R. R. Gaunt to R. Paruolo, " Millstone Nuclear Power Station - Unit 3
~ Update - Required Instrumentation for Station Blackout (SBO) Recovery Plan," GIC 89-130, dated March 29, 1989.
17.
Northeast Utilities Inter Office Memorandum from R. C. Thomas to G. E. Cornelious, " Quality Assurance Req'airements For Station Blackout (SBO) All Units," GMB-89 R.441, dated August 11, 1989.
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18.
- Mroczka, E.'J., letter to T. E. Murley (NRC), "Haddam Neck Plant, Millstone Nuclear Power Station Units Nos. 1, 2, and 3,. Response to l
- Station Blackout, Additional information," Docket Nos. 50 213, 50-245, 50 336, 50-423, dated March 30, 1990.
19.
Thadani, A. C., letter with attachment to W. H. Rasin of NUMARC,
" Approval lof NUMARC Documents on Station Blackout (TAC-40577)," dated j
October 7 1988.
i 20.-
Thadani, 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).
21.
Northeast Utilities Inter Office Memorandum from Gaunt, R. R. and Stricker, S. I, to C. Ashton, " Millstone. Unit 3 - Station Blackout
~
(SBO), Heat Loads For Main Control Room I&C lighting during SBO," GIC-l 89-110, dated March 10, 1989.
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