ML20085C172
| ML20085C172 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/26/1991 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20085C173 | List: |
| References | |
| CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-91-6662, TAC-68555, NUDOCS 9109030085 | |
| Download: ML20085C172 (37) | |
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Attachment 1 SAIC-91/6662 TECHNICAL EVALUATION REPORT HOPE CR EK GENERAT1HG STATION STATION BLACKOUT EVALUATION TAC No. 68555
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" EE awa Sc>ence Applications kstematmnalC01pa &ilaci An EmployeeDwned Company Final August 26, 1991 b
Prepared for: (
U.S. Nuclear Regulatory Comission Washington, D.C. 20555 ,
Contract NRC-03-87-029 Task Order No. 38 1710 Goodndge Drive PO. Box 1303. McLean. Vrgonna 22102 I?O3) 821 4300
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TABLE OF CONTENTS Section .P_iLqq
1.0 BACKGROUND
........................................... 1 2.0 REVIEW PROCESS ....................................... 3 3.0 EVALUATION ........................................... 6 3.1 Proposed Station Blackout Duration ............. 6 3.2 Station Bl ackout Coping Capability . . . . . . . . . . . . . 10 3.3 Proposed Procedures and Training ............... 23 3.4 Proposed Modifications ......................... 24 3.5 Quality Assurance and Technical Specifications . 25
4.0 CONCLUSION
S .... ..................................... 26
5.0 REFERENCES
........................................... 32 11 3
TECHNICAL. EVALUATION REPORT HOPE CREEK GENERATING STATION STATION BLACKOUT EVALUATION
1.0 BACKGROUND
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 l Alternating Current Power" (1). The objective of this requirement is to assure that all nuclear power plants are capable of withstanding a station blackout (580) and maintaining adequate reactor core cooling and appropriate containment integrity for a required duration. This requirement is based on informatioa 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 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 l
guidance for addressing the 10 CFR 50.63 requirements. The application of l
this method results in selecting a minimum acceptable SB0 duration capability from two to sixteen hours depending on the plant's characteristics and l
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 einergency i
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-revicw prc;ess,-NUMARC developed two-generic response documents. These documents were reviewed and endorsed (9) by the NRC staff for the purposes of plant specific submittals. The documents are titled:
- 1. " Generic Response to Station Blackout Rule for Plants Using Alternate AC Power," and
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- 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 i summary of results of the analysis of the plant's station blackout coping capability. Licensees are expected to ensure that the baseline assumptions used in NUMARC 87-00 are applicable to their plants and to verify the accuracy of the stated results. Compliance with the 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 implemer.ted the necessary changes as required to meet the SB0' rule.
In 1989, a joint NRC/SAIC team headed by an NRC staff member performed audit reviews of the methodology and documentation that support the licensees' submittals for several plants. These audits revealed several deficiencies which were not apparent from the review-of the licensees' submittals using the agreed upon generic response format. These deficiencies raised a generic
-question regarding the degree of licensees' conformance to the requirements of the SB0 rule. To resolve this question, on January 4, 1990, NUMARC issued -
additional guidance as NUMARC 87-00 Supplemental Questions / Answers (10) addressing the NRC's concerns regarding the deficiencies. NUMARC requested that the licensees send their supplemental responses to the NRC addressing; d
these concerns by March 30, 1990.
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'2.0 REVIEW PROCESS The review.of the licensee's submittal it. focused on the following areas consistent with the positions of RG 1.155:
A. Minimum acceptable SB0 duration (Section 3.1),
B. SB0 coping capability (Section 3.2),
C. Procedures and training for SB0 (Section 3.4),
D. Proposed modifications (Section 3.3), and E. Quality assurance and technical specifications for 5B0 equipment (Section 3.5).
For the determination of the proposed minimum acceptable SB0 duration. l the following. factors in the licensee's submittal are reviewed: a)-offsite ,
power design characteristics, b) emergency AC power system configuration, c).
determination of tha emergency diesel generator (EDG) reliability consistent
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with NSAC-108 criteria (11), 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 l'.155 provides a matrix for determining-the required coping duration.
For the 580 coping capability, the licensee's submittal-is reviewed to i assess the availability, adequacy and capability of the plant 1 systems and components needed to achieve and maintain a safe shutdown condition and recover from an SB0 of acceptable duration which is determined above. The review process follows the guidelines given in RG 1.155, Section 3.2, to assure:
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- a. availability of suff tetent condensate inventory for decay-heat-removal, 3
- b. adequacy of the class-lE 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 pl ant,
- 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 (i.e., revised existing and new) for coping with 580 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, ventilation systems, containment isolation valves, and primary coolant make-up capability is reviewed. Technical specifications and quality assurance set forth_by the licensee to ensure high reliability of_the' equipment, .
specifically added or assigned to meet the requirements of the SB0 rule, are assessed- for their adequacy.
The licensee was asked to-respond to a set of questions dated November 13, 1990, forwarded to the licensee by the NRC. In response to these questions and a request for a telephone conference to resolve the licensee's l
response to the SB0 rule, the licensee said that it would not be ready to discuss its submittals until February 28, 1991. On March 1, 1991, the licensee sent a letter stating that it would provide information:by March 28, 1991. On March 28, 1991, the licensee _provided additional information on its
__re analysis of its 580 coping ability. The licensee, however, did not- .
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9 specifically address the questions. Consequently, this SB0 evaluat1on is based upon the review of the licensee's submittals dated April 17, 1969 (12),
April 30, 1990 (13), June 26, 1990 (15), July 30, 1990 (16), March 1, 1991 (17), and March 28,'1991 (18), and the information available in the plant Updated Final Safety Analysis Report (UFSAR) (14). An audit may be warranted as an additional confirmatory action. This determination would be made and the audit would be scheduled and performed by the NRC staff at scme later date.
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3.0 EVALUATION Our review is based upon the licensee's submittals and the information available in the UFSAR. The licensee was asL!' to respond to a set of questions on November 13, 1990, concerning its submittals in the areas of ESW grouping, battery-capacity calculations, loss of HVAC, and the conditions of the reactor vessel at the end of the $80 event. The licensee indicated that it was in the process of revising its submittals and would provide its revised submittal by February 28, 1991. After receiving the licensee's submittal dated March 28, 1991, the licensee was asked to arrange a telephone call to resolve the concerns outlined in the sct of questions forwarded to the licensee earlier. However, the licensee declined to beve a telephone conversation expeditiously. Since the review of the plant 580 submittal has been delayed considerably by the licensee's failure to respond, the decision was made to review the plant coping capability based on the availsble information contained in the submittals and the plant UFSAR. It should be noted that, per the SB0 rule, the licensee should have finished its ccping evaluation by April 17, 1989, and be prepared for any follow up inspection and audit of its supporting analysis. The licensee's inability to respond to the questions for which answers should have already been provided indicates the lack of having a completed analysis by March 30, 1990, as dictated by the SB0 rule and follow-up guidance.
3.1 Proposed Station Blackout Duration Licensee's Submittal The licensee, Public Service Electric and Gas Company (PSE&G),
calculated (12 and 13) a minimum acceptable station blackout duration of four hours for the Hope Creek Generating Station (HCGS). The licensee stated that no modifications are required to attain this coping duration.
The plant factors used to estimate the. proposed SB0 duration are:
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- 1. Offsite Power Design Characteristics The plant AC power design characteristic group is "Pl" based on:
- a. Independence of the plant offsite powar system characteristics of "I1/2,"
- b. Expected frequency of grid-related LOOPS of less than one per 20 years,
- c. Estimatcd frequency of LOOPS due to extremely severe weather (ESW) which places the plant in ESW Group "2," and
- 2. Emergency AC (EAC) Power Configuration Group The EAC power configuration of the plant is "C." HCGS is equipped with four emergency diesel generators, two of which are necessary to operate s'afe-shutdown equipment following a loss of offsite ;
power. Of the four EDGs (A, B, C, and D), either A and C or B and D are necessary to safely shut down the plant.
- 3. Target Emergency Diesel Generator (EDG) Reliability The licensee selected a target EDG reliability of 0.95. The selection of this target reliability is based on having a unit average EDG reliability of greater than 0.90 for the last 20 demands consistent with NUMARC 87-00, Section 3.2.4. The licensee statea (18) that since HCGS began operating in December, 1980, there have been 81 demands for EDGs A and B, and 79 for EDGs C and D. During this period, the average unit EDG reliability has been -
0.9935.
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The licensee initially stated (13) tt.. there is a broad range of independent activities which are conducted to assure that EDG reliability is maintained at an acceptable level, and that it ~ is following NUMARC and NRC activities regarding resolution of Generic Issue B-56. -Th( 'icensee added that upon resolution of B-56, it will formalize the reliability program including the use of the EDG target reliability value. In its revised submittal (18),
the licensee stated that there are existing surveillance terting and performance-monitoring procedures designed to track EDG performance and support maintenance activities.
Review of Licensee's Submittal Factors which affect the estimation of the S80 coping duration are: the independence of the offsite power system grouping, the estimated frequency of LOOPS due to ESW and SW conditions, the ex,pected frequency of grid-related LOOPS, the classification of EAC, and the selection of
-EDG target reliability. Using Table 3-3 of NUMARC 87 00, the expected frequency of LOOPS at HCGS due to SW condition is group "2," which is censistent with that given in NUMARC 87-00. Using Table 3-2 of NUMARC 87-00, the expected frequency of LOOPS due to ESW conditions place the HCGS site in ESW group "4." In its submittal, the licensee stated that if site-specific data is used, its ESW group is "2." This change in ESW classification places the site in an offsite power design characteristic group "P1," inst ead of "P2" if ESW group "4" is used. Since the licensee failed to justify the discrepancy between its ESW grouping and the one provided in NUMARC 87-00 as was requested earlier,-we consider the site to be classified as that given_in NUMARC 87-00 for ESW grouping.
The licensee stated that the independence of the plant offsite power system grouping is "I1/2." A review of the HCGS UFSAR indicates that:
- 1. All power sources come to the site through a single' switchyard; l - _ __
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- 2. There are two station service transformers, each connected to two of the four emergency buses;
- 3. Etch transformer is sized and designed to support both divisions; and
- 4. Upon loss of either offsite power source, the remaining transformer can power beth divisions (all four emergency buses) through an automatic transfer.
Based on these and the criteria stated in Table 5 cf RG 1.155, we conclude that the plant independence of offsite power system group is "I2."
The licensee correctly classified the EAC configuration of HCGS as "C."
At Hope Creek, there are four diesel generators, two of which are necessary to safely shut down th e pl ant. However, not all combinations of two EDGs pcwer the necessary equipment. The tour EDGs are labelled A, B, C, and D. Either A and C or B and D are necessary to power one mechanical division to shut down the plant. For this analysis, we assumed that A and C are one EDG, and B and D are another EDG, making Hope Creek have the equivalent of two EDGs, any one of which is necessary to shut down the plant.
Based upon the information given in the licensee's submittal, it appears that the EDG target reliability the licensee selected (12) and committed to maintain (13) is appropriate. The licensee selected an EDG target reliability of 0.95 based on the EDG reliability data for the last 20 demands. The licensee noted that the EDG reliability to date is 0.9935.
Based on the data provided, it appears that there has been one EDG failure in the last 160 demands on the four EDGs. However, we are unable to verify the EDG reliability assessment, since the EDG failure statistics will only be available on site for review. Based on the reported EDG failure statistics, the licensee can choose any EDG target i reliability consistent with the minimum required SB0 durationc l
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i Although the licensee has existing EOG testing and performance-monitoring procedures, it did not state how the procedures conform to the guidance provided in RG 1.155, Appendix 0 to NUHARC 87-00, or in the esolution of the Genet ic fafety Issue B 56. The licensee needs to verify that its EDG reli bility program conforms to the program outlined in the guidance.
With regard to the expected frequency of grid related LOOPS at the site, we can not confirm the st3ted results. The availabl( information in NUREG/CR 3992 (3), which gives a compendium of information on the loss of of fsite power as nuclear power plants in the U.S., only provided infermation through the calendar year 1984. '9 wever, HCGS did not enter commercial operation until 1986. In the abt ts of any contradictory information, we agree with the licensee's sta ement.
Based on the ESW and SW groepings provided in NUMARC 87-00, HCGS is in power design characteristic prvip "P2," and, in order for the plant to be a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping site, the lic^nseo needs to commit to a target EDG reliability of 0.975 instead of its present 0.95. Selection of an EDG target reliability of 0.95 requires the licensee to resubmit the plant coping capability analysis for an eight-hour duration. Our review of the plant coping capability is based on the licensee changing the target EDG reliability to 0.975, thereby requiring the plant to cooe for four hours.
3.2 Station Blackout Coping capability The plant coping capability with an 5B0 event for the required duration of four hours is assessed with the following results:
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1, . Condensate Inventory fcr Decay-Heat Removal Licensee's submittal l
1he licensee initially stated (12) that 91,000 gallons of water are reautred for decay-heat remaval and cooldown for four hours !
and that the minimum permissible condensate storage tank (CST) level per plant Technical Specifications corresponds to 135,000 gallons of water. The licensee stated (13) that the condensate-inventory calculations reautre re review to ensure that all reactor coolant leakage losses are accounted for and that the plant specific analysis performed in suppc*' a primary system coolant inventory assessment is consisten6 ritn the conditions existing during an SB0 event.
In its revised submittal (18), the licensee stated that 132,606 gallons of water are required to cope with a four hour 580 event.
This value includes the condensate necessary for decay heat removal (72,841 gallons), cooldown (24,500 gallons), level shrinkage (19,425 gallons), and leakage (15,840 gc11ons). The licensee concluded that it has sufficient water to cope with a four-hour SB0 event.
Review of Licensee's Submittal Using the expression provided in NUMARC 87-00, we have estimated that 74,298 gallons of water would be required to remove decay heat during a four-hour SB0 event. This estimate is based on 102%
of a maxim:;m licensed core thermal ra'ing of 3293 MWt. In adaition, condensate has to be provided to account for an assumed ieak rate of 66 gpm (18 gpm per pum;: and an allowed technical specification leak ratt of 30 gpm), which is 15,840 gallors over four hours.' Therefore, a total'of 90,138 gallons of condensate would be necessary for decay heat removal and to compensate for leakage, 11
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Since the licensee plans to cool down the plant, additional water [
will he needed. The licensee, however, did not provide the final conditions of the plant at the end of the four hour SB0 event, so we are unable to determine the amount of the water necessary for f cooldown.
The licensee stated that the plant Technical Specifications l requires a' minimum CST level corresponding to 135,000 gallons of condensate be available, leaving 44,862 gallons for any necessary l cooldown. In the questions sent to the licensee, we asked about the plant conditions at the end of the 5B0 event. The licensee !
. did not provide any information on either the reactor-vessel ;
conditions or the suppression-pool temperature at the end of the SB0 event. Based on our analyses of similar plants, we concluded the remaining 44,862 gallons is sufficient for any necessary il cooldown. The licensee, however, needs to ensure that the technical specification minimum CST level is applicable during plant operation. ;
- 2. Class-1E Bat.tery Capacity t.icensee's Sutelttal The licensee stated (12) that a battery capacity calculation was performed to verify that the class-1E batteries have sufficient capacity to meet station blackout loads for four hours. The licansee stated that the plant 125 and 250-VDC battery-load studies which identify the battery loads for the most severe plant l design basis accident condition of coincident LOOP and loss of-coolar,t accident (LOCA) have been used as the basis for determining the battery SB0 duty cycle. The licensee-added (18) that.the dut.y cycle includes the AC power re5toration loads at the -
most limiting time in the battery duty cycle and-do not include any manual load stripping. ,
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d For the calculations, the licensee followed IEEE Std 485, using a minimum electrolyte temperature of 72*F, a design margin of 1.1, and an aging factor of 1.25. The licensee stated that the minimum battery terminal voltage reachect during the bat.tery SB0 duty cycle is 105 and 210 VDC for the 125 and 250-VDC systems, respectively.
The licensee determined that there are two batteries (125 VDC batteries IAD411 and ICD 411) which have no excess irargin over the 1.1 design margin when a minimum terminal voltage of 105 VDC is used. The licensee also determined that the other 125. and 250-VDC batteries have available excess capacity over the 1.1 design margin. The licensee concluded (18) that the results of the battery capacity evaluation verify that the existing HCGS battertes have adequate capacity to supply their connected SB0 33 ads for the 4-hour coping duration.
Review of Licensee's Submittal At HCGS, there are both 125 and 250-VOC class lE Sattery systems.
The class-lE 125-VOC system is divided into four independent channel systems. Each of these systems is supplied from batteries of the corresponding load group channel. There are two 250-VOC .
systems. One is associated with the high pressure coolant injection (HPCI) system, and the other is associated with the reactor core isolation cooling (RCIC) system.
Our review consists of examining the battery loads, Tables 8.3-7 through 8.3-10, and the load profiles, figure 8.3-16, given in the plant UFSAR, and the licensee's submittals. Upon reytew of this information, we believe that the loads are consis at with that which would be expected during an SB0 event. Based on these and the information given in Section 8.3.2.1.2.2 of the HCGS UFSAR, it appears each battery bank has sufficient capacity to indeperdently supply the required loads for four hours without support from the battery chargers.
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From the information provided by the licensee, we found only one ;
non conservative assumption. The licensee's assumption of a l minimum electrolyte temperature of 72*F is non conservative. The use of this temperature is satisfactory if the licensee can ensure that, under all circumstances, the temperature in the battery room will remain at or above 72'F.
Based on the information available in the UFSAR and that provided by the licensee, we have the following concerns with the battery calculations:
- 1) It appears that the licensee assumed only one attempt at starting the EDGs. The licensee needs to verify that, should the operators attempt to start the EDGs more than once, there will be sufficient battery capacity. -
- 2) The licensee assumed a final terminal voltage of 105 and 210 -
VOC for the 125- and 250 VDC batteries. The licensee needs to verify that all of the 580 equipment can be operated at the assumed minimum battery terminal voltage.
- 3) If the electrolyte temperature has a low-temperature alarm in the control room, then, at the initiation of the 580 event, the minimum electrolyte temperature will be 72'F..
However, if the SB0 avent occurs when the outside temperature is the minimum expected and there is no heat available in the battery room, there is the possibility that the battery-room temperature will drop Wer the course of the SB0 event, resulting in a battery capacity which is less than that which the licensee calculated. Therefore, the licensee needs to evaluate the battery capacity for the condition when there is no heat available in the battery room an,d the outside temperature is the minimum expected.
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- 4) The 125 VDC batteries lA0411 and 100411 have essentially zero margin. Therefore, any future load addition requires a re analysis of the battery capacity.
- 3. Compressed Air Licensee's Submittal The licensee stated that the air operated valves relied upon to cope with an SB0 for four hours have sufficient back up sources independent of the class-lE power supplies. The licensee also stated that the Safety Relief Valves (SRV) which will be used for ,
plant depressurization are accumulator-backed with sufficient capacity to operate for the four-hour SB0 cvent.
Review of Licensee's Submittal At Hope Creek, the SRV will be used for depressurization during an SB0 event. Each SRV used for this purpose is provided with an accumulator capable of opening the valve twice against 70% of the maximum drywell pressure of 70 psig. Therefore, HCGS has sufficient back up air supplies to cope with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO.
- 4. Effects of loss of Ventilation Licensee's Submittal The licensee stated (12) that calculations were performed to determine the steady-state ambient air temperature in dominant areas of concern following the loss of the Heating Ventilation, and Air Conditioning (HVAC) system. The licensee initially stated (12) that the peak temperatures in the high-pressure coolant injection (HPCI) pump room and the reactor core isolation cooling (RCIC) room would be 151 F and 138*F, respectively.
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The licensee added (13) that it used a combination of vendor-specific analysis and NUMARC 87-00 methods to determine dominant areas of concern and that a review of this combination is needed to ensure that a conservative result was obtained. The licensee noted that any changes to results obtained in the original review will need to be assessed for equipment operability.
In its revised submittal (18), the licensee stated that the average steady state temperature in dominant areas of concern containing equipment necessary to achieve and maintain safe shutdown during an SBC event has been calculated using the assumptions provided in NUKARC 87-00, Section 2.7.1 and the methodology provided in NUMARC 87-00, Section 7.2.4 and Appendix E. The licensee also stated that credit has been taken for opening area doors where feasible during the SB0 event to allow for removal of heat through natural circulation. The licensee calculated the effects of opening the doors per the methods provided in NUMARC 87 00, Section 7.2.4, and/or Appendix E. The licensee provided (18) the following results of these calculations:
Closed Open Room Doors ('F) Doors ('F)
RCIC Room 134.8 HPCI Room 173.2 Battery Room 5104 94.3 Battery Room 5128 93.6 Control Equip. Room 157.3 142.0
- App. E Approach 134.5
- Reduced Heat Load 118.3 Inverter Room 5447 105.8 Inverter Room 5448 127.3
, Elec. Access Room 120.0 Control Room With Acoustic Ceiling 176.6
- Without Acoustic Ceiling 120.0 :120.0 Inverter Room 5615 129.4 111.7 16
4 The licensee stated (18) that an analysis has been made for i control. room habitability during the SB0 event based upon control-room temperature and humidity. The results of this analysis indicate that the control room habitability will not be affected v; ring the SBO. The control-room and lower control equipment room ;
temperatures were calculated to be less than 120'F based upon the opening of the control room and lower control equipment room doors. The licensee added (12) that provisions for the opening of these doors are being incorporated into plant procedures. The licensee noted that the upper control equipment room temperature will not exceed 120'F.
The licensee also stated (12) that the temperature in the main steam tunnel was not evaluated because there is no equipment required to function during a statico blackout there; therefore, the main steam tunnel is not a dominar.t area of concern (DAC).
The licensee stated that reasonable assurance of the operability of SB0 response equipment in the above OACs has been assessed using Appendix F to NUMARC 87 00 and the Topical Report, and determined that no modifications are required to provide reasonable assurance of equipment operability.
The licensee stated (13) that it is reassessing its drywell and suppression-pool heat-up calculations. The computer model used for assessing drywell and suppression-pool heat-up may not be accurate in certain areas, in that it utilized a General Electric Mark 11 containment and suppression pool in lieu of a Mark I containment tnd torus. The licensee stated (13) that this analysis requires further review and justification as to its validity.
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I Review of Licensee's submittal The licensee provided (18) information on the final calculated temperatures in the areas of concern. We did not, however, receive information on the licensee's assumptions for initial temperatures or heat loads and, therefore, we are unable to perform a review of the licensee's calculations. The licensee )
needs to provide and verify that the initial temperatures are the maximum allowable and that the heat loads accurately reflect those ;
expected during an SB0 event.
With regard to the control-room calculation, the licensee did not explain what it means by "with acoustic ceiling" and "without acoustic ceiling." We do not know if this means that the licenses plans to make a modification to the control room to permanently -
remove the acoustic ceiling or whether it means that the operators will remove part of the acoustic ceiling during an SB0 event, if permanent changes to the acoustic ceiling are required, the licensee needs to identify the modification in its SB0 submittals 1 supporting documentation. If temporary changes to the acoustic ceiling are necessary only durir.g an SB0 event, the. licensee needs ,
to state how it determined the number of acoustic ceiling tiles, and the locations from where they-should be removed, in order to provide additional surface area for the heat sink in the control I room. If it is necessary to remove ceiling tiles, the licensee needs to establish a procedure in which guidance is provided for -
the location of the tiles to be removed and the time at which they must be removed.
In the calculation for the control equipment room temperature, the licensee calculated three different temperatures with the doors open. It~is unclear what the differences are between these three
! temperatur'es, what assumptions were used to calculate them, and-l which temperature the licensee considered when performing an
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1 1ssessment of the retsonable assurance of equipment operability in this room.
With regard to the calculated final temperatures, in several cases the licensee took credit for the opening of doors. The licensee, however, did not state whether it is planning to revise any nrocedures for instructing the operators to do so within 30 inutes of the onset of the SB0 event, as guided in NUHARC 87 00.
We have no information concerning the licensee's re-assessment of the drywell or suppression pool heat up calculations. In general, there is a concern about the drywell heat up in a Mark I containment during an SB0 event. Even though the licensee is planning to cooldown the reactor, there is the potential for high temperatures in areas where equipment needs to be operational.
Therefore, we believe that it is necessaiy for the licensee to verify that all assumptions made for its drywell calculation are conservative and that the assumed initial conditions accurately reflect those expected during an SB0 event.
The licensee dio not perform a calculation for the main steam tunnel because it stated that there is no equipment which is required to be operable. The lack of a temperature evaluation in the main steam tuur.el is only accepthble if 'no equicment needs to be operational during an SB0, and all equipment (i.e., valves or other (quipment) are in safe positions. For example, all valves are fully closea automatically at the beginning of the event or are closed within a short period of the time of the incident, before there is an appreciable hedt-up. If the licensee verifies that there are valles in the main steam tunnel which need to be operable should containment isolation be necessary, the licensee needs to perform a heat up calculation and assess the operability of thete valves in order to ensure the capability to provide.
adequate containment integrity, or provide justification (s) as to 19
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why the valves will remain operational at the highest expected [
steam tunnel temperature. ,
- 5. Containment Isolation !
Licensee's submittal l i
.The licensee stated that the plant list of containment isolation l valves (CIVs) was reviewed and it was determined that all of the valves wi.tch must be capable of being closed or operated (cycled) under SB0 conditions can be positioned with indication independent of the preferred power supplies. The licensee also stated that those CIVs not required to operate in response to the event are assumed to be in their normal position upon loss of power and that ;
positfua verificatica for these vtives is not required.
In addition-to the evitaria outlined in RG 1.155, the licensee :
nsed (18) two additional exclusion criteria:
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- 1) Water seals: Suction inlets and discharge points are ;
submerged below the water level in the suppression pool. l The water-in the pipe provides a barrier from the containment atmosphere.
- 2) Valves that must be closed for reactor operation: The RHR shutdown cooling suction valves and the RHR head-spray valves are interlocked closed by reactor high pressure signal. The RHR containment spray valves are interlocked closed on reactor level and drywell pressure signals. -These valves must be closed for proper reactor operation.
i The licensee stated (18) that, based upon its evaluation of containment isolation capability during an SB0 event, all-valves which require closure capability can be closed or verified closed by manual operator actions. The licensee initially stated (12)
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that modifications to the inboard main steam isolation valve (MSIV) drain line, drywell sump drain line, and drywell equipment drain line are rr uired in order to ensure that, when appropriate, containment integrity can be established. However, in its most recent submittal, the licensee concluded (18) that no plant modification is required to comply with the requirements of 10 CFR 50.63.
Review of Licensee's Submittal Upon review of UFSAR Table 6.2-16, we found that there are several valves (i.e.. core spray pump suction, residual heat removal shutdown cooling suction, etc.) which do not meet the exclusion criteria outlined in RG 1.155. The licensee excluded (18) valves which must be closed for proper reactor operation and those that are on suction or discharge lines which terminate below the suppression-pool water level.
The additional exclusion criteria assumed by the licensee are not consistent with those outlined in RG 1.155:
- 1) Although the valves which are on lines that terminate below the suppression pool surface were granted exemption from testing for leak tightness under 10 CFR 50 Appendix J, this cannot be construed as an exclusion criterion for isolation capability requirements. For this reason, those valves which were excluded by this provision need to be treated as other CIVs not excluded by the criteria allowed by RG 1.155.
- 2) Valves which must be closed for proper reactor operation cannot be excluded unless they are " locked" closed.
Therefore, the closure of these valves needs to be verified.
It is unclear whether or not the licensee plans to make plant modifications to several valves which were listed in the. submittal .
21
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dated April 17, 1999 (12). The licensee's most recent statement (18) indicates that no modifications are necessary in order to ensure that proper containment integrity can be provided during an SB0 event.
The assurance of adequate containment integrity requires that the operators be aware of the CIVs positions at all times. Since during an SB0 the AC-operated valves will not have position indication in the control room, the licensee needs to list in an appropriate procedure the CIVs which are either normally closed or open and fail as-is upon loss of AC power and cannot be excluded by the criteria given in RG 1.155, and identify the actions necessary to ensure that these valves are fully closed, if needed.
Valve closure needs to be confirmed by position indication (local, mechanical, remote, process information, etc.).
- 6. Reactor Coolant Inventory Licensee's Submittal The licensee stated (12) that the ability to maintain adequate reactor coolant system inventory to ensure that the core is cooled has been assessed for four hours. The licensee stated (10) that a reactor coolant inventory analysis has been performed in accordance with NUMARC 87-00, Section 2.5, using KAAP computer codes and Hope Creek plant-specific models, and these dynamic models also demonstrate that adequate reactor coolant inventory is maintained for a coping duration of four hours and that the core stays covered.
The licensee concluded (12) that the HPCI and RCIC pumps, which take suction from the CST, provide the necessary flow to maintain adequate reactor coolant system (RCS) inventory to ensure that core uncovery does not occur during the require coping duration.
The licensee stated that make-up systems in addition to those ,
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available under SB0 conditions are not required to maintain adequate core cooling.
Review of Licensee's Submittal Reactor coolant make up is necessary to remove decay heat, to replenish the RCS inventory losses due to leakage, and to provide water for any cooldown attempted during the four-hour SB0 event.
In order to estimate the average condensate injection rate necessary to keep the core covered, we assumed that the entire volume of the CST (135,000 gallons) is depleted af ter four hours.
This yields an average injection rate of -560 gpm. The RCIC pump has an injection rate of 600 gpm and the high-pressure coolant injection (Hr",1) pump has an injectica rate of 5600 gpm. Both of these pumps are turbine-driven and take suction from the CST.
Therefore, Hope Creek RCIC and HPCI pumps have su.fficient capacity to compensate for the average expected inventory-loss rate for four hours.
NOTE:
The 18 oom recirculation Dumo seal leak rate was agreed to between NUHARC and the NRC staff pending resolution of Generic issue (GI) 23. If the final resolution of GI-23 defines higher recirculation pump seal leak rates than assumed for the RCS inventory evaluation, the licensee needs to be aware of the potential impact of this resolution on its analyses and actions addressing conformance to the SB0 rule.
3.3 Proposed Procedures and Training Licensee's Submittal The licensee stated that the following plant procedures have been reviewed per guidelines in NUMARC 87-00, Section 4:
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- 1. Station blackout response glidelines,
- 2. AC power restoration, and
- 3. Severe weather.
The licensee stated that these procedures have been reviewed and the changes necessary to meet NUMARC 87 00 guidelines will be implemented.
Review of Licensee's Submittal We neither received nor reviewed the.affected SB0 procedures. These procedures are plant specific actions concerning the required activities to cope with an 580. It is the licensee's responsibility to revise and.
implement these procedures, as needed, to mitigate an SB0 event and to assure that these procedures are complete and correct, and that the associated training needs are carried out accordingly.
3.4 Proposed Modifications Licensee's Submittal The licensee initially stated (12) that modifications to the inboard MSIV drain line, drywell sump drain line, and drywell equipment drain line are required in order to ensure that containment integrity can be established, in a later submittal (18), the licensee stated that no-modification is required to comply with the requirements of 10 CFR 50.63.
Review of Licensee's Submittal We did not receive any information on why the initial proposed l
l' modificatio.n was needed, nor did we receive any explanation of the
! reversal in decision that there is no need for any modifications. The licensee needs to explain the original decision and the reason (s) for its reversal.
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As part of its control room heat up calculations, the licensee proposed to remove control room acoustic ceiling tiles during an SB0 event in order to provide an adequate heat sink to keep the control-room temperature below 120'F. The licensee needs to provide information on when and from where the tiles will be removed. Additionally, our review has identified several concerns which may require modifications for their resolution.
3.5 Quality Assurance and Technical Specifications The licensee did not provide documentation on how the plant complies with the requirement of RG 1.155, Appendix A.
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4.0 CONCLUSION
S Based on our review of the licensee's submittals and the information ,
available in the UFSAR for Hope Creek Generating Station, we find that the submittal conforms with the requirements of the 5B0 rule by following the <
guidance of RG 1.155 with the following exceptions:
- 1. Offsite Power Characteristics
- a. Extremely Severe Weather (ESW) Grouoina Using Table 3-2 of NUMARC 87-00, the expected frequency of LOOPS due to ESW conditions place the HCGS site in ESW group <
"4." In its submittal, the licensee stated that if site-specific data is used, its ESW group is "2.* This change in ESW classification places the site in an offsite power design characteristic group "P1," instead of "P2" if ESW group "4" is used. Slece the licensee failed to justify the ,
discrepancy between its ESW grouping and the one provided in NUMARC 87-00 as was requested earlier, we consider the- site to be classified as those given in NUMARC 87 00 for ESW and SW groupings.
- b. Taraet EDG Reliability The licensee needs to change the target EDG reliability from 0.95 to 0.975 in order to be a 4-hour coping plant in stead of an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> t jing plant. This change is necessary because the licensee used an ESW grouping different from that given ,
in NUMARC and failed.to adequately justify its use. The licensee also did not state how its EDG reliability procedures conforms to the guidance provided in RG 1.155 and
! NUMARC 87-00, Appendix 0. <
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e A
- 2. Condensate :nventory The licensee needs to ensure that the plant Technical Specification for the minimum CST water level corresponding to 135,000 gallons of water is applicable during plant operations.
.3. Class-lE Battery capacity ;
Although it appears that the batteries are designed to last for- ,
four hours, the licensee needs to respond to the following items:
- 1) It appears that the licensee assumed only one attempt at starting the EDGs. The licensee needs to verify that, should the operators attempt to-start the EDGs more than once, there will be sufficient battery capacity. ,
- 2) Th4 dicensee assumed a final terminal voltage of 105 and 210 -
VDC for the 125- and 250 VDC batteries. The licensee needs j to verify that all of the SB0 equipment can be operated at the assumed m'..imum battery terminal voltage.
- 3) Th3 licensee needs to vvaluate the battery capacity for the condition when there is no heat available in the battery room and the outside-temperature is the minimum expected.
During an extremely cold condition, there is a potential for the battery-room temperature to be below the temperature for which the licensee evaluated the battery capacity, which
]
could result in the battery capacity being less than that i which the licensee calculated. ,
- 4) The 125 VDC batteries IA0411 and 100411 have essentially-zero margin. Therefore, any future loao addition requires a re-analysis-of the battery capacity.
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- 5) The licensee's assumption of a minimt.m electrolyte timperature of 72'F is non conservative. The use of this temperature is satisfactory if the licensee can ensure that, under all circumstances, the temperature in the battery room will remain at or above 72'F. l I
- 4. Effects of Loss of Ventilation
- a. Initial Temoprature Assumotig_nj n The licensee did not provide any information :n its assumptions for initial temperatures or heat loads and, therefore, we are unable to perform a review of the licensee's calculations.
- b. Control Room Heat-Up With regard to the control-room calculation., the licensee did not explain what it means by "with acoustic ceiling" and "without acoustic ceiling." If this means that permanent changes to the acoustic ceiling are required, the licensee neods to identify the modification in its SB0 submittals ;
supporting documentation If it means that-temporary changes to the acoustic ceiling are necessary only during an SB0 event, the licensee needs to state how it datermined how many ceiling tiles, and from where it should remove the acoustic tiles, in order to provide sufficient cooling in the control room.
.- c. Control Room Eauipment M In the calculation for the control equipment room temparature, the licensee calculated three different temperatures with-the doors open. It is unclear what the differences are between these three temperatures and which is the temperature the licensee considered when assessing .
l the reasonable assurance of equipment operability for the room.
28 c,--- ,,-,1, -n--n-. - . - .,e.we- -n--, -----,,,,,-e
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- d. Procedure for Ooenina Doors With regard to the calculated final temperatures, in several cases the licensee took credit for the opening of doors.
The licensee, however, did not s nte whether the opening of these doors is being incorporated into the plant SB0 procedures, and therefore needs to ensure that the operators are instructed in an appropriate procedure to open the doors.
- e. Drywell H, eat-tJE We have no information concerning the licensee's re-assessment of the drywell or suppression pool heat-up calculations. The licensee needs to verify that all assumptions made for its drywell calculation are, conservative and that the assumed initial conditions accurately reflect those expected during an SB0 event,
- f. Main Steam Tunnel The licensee did not perform a calculation for the main steam tunnel because it stated that there is no equipment which is required to be operable, if there are valves in the mairi steam tunnel which need to be operable should containment isolation be necessary, the licensee needs to perform a heat-up calculation and assess the operability of these valves.
- g. Reasonable Assurance Of Eouipment Ooerability The licensee needs to verify that the SB0 equipment in the following rooms will be operational following the loss of HVAC:
- Control Equipment Room
- Inverter Rooms 5447, 5448, and 5615 Cont.rol Room.
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These rooms were chosen because the ambient air temperature is greater than 104*F. In general, the temperature inside cabinets is 15 20'F higher than the bvik air temperature in the room. Taking this into consider, tion, the licensee needs to verify that there is reas',nable assurance of equipment operability in these rLoms, especially in the cabinets.
- 5. Containment Isolation Upon review of UFSAR Table 6.2-16, we found that there are several valves (i.e., core spray pump suction, residual heat removal shutdown cooling suction, etc.) which do not meet the exclusion criteria outlined in RG 1.155. The licensee excluded (18) valves which must be closed for proper reactor operation and thosn that are on suction or discharge lines which terminate below the suppression pool water level.
- 6. Proposed Modifications The licensee initially stated (12) that modifications to the inboard MSIV drain line, drywell sump drain line, and drywell equipment drain line are' required in order to ensure that containment integrity can be established. In a later submittal (18), the licensee stated that no modification is required to comply with the requirements of 10 CFR 50.63. The licensee needs to have a clarification to this discrepancy in its 5B0 submittals supporting documentation.
As part of its control room heat-up calculations, the licensee proposed to remove control-room acoustic ceiling tiles during an SB0 event in order to provide an adequate heat sink to keep the control-room temperature below 120*F. The licensee needs to provide information on when and from where the tiles will be 30
removed. Additionally, our review has identified several concerns which may require modifications for their resolution.
- 7. Quality Assurance and Technical Specifications The licensee's submittals do not provide any information on how the plant's SB0 equipment conforms with the guidance of RG 1.155, Appendix A.
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5.0 REFERENCES
1.
The Office of Federal Reoister, " Code of Federal Regulations Title 10 Part 50.63," 10 CFR 50.c3, January 1, 1989.
2.
U.S. Nuclear Regulatory Commission, " Evaluation of Station Blackout Accidents at Nuclear Power Plants - Technical Findings Related to W., June 1988.
Unresolved Safety Issue A 44," NUREG-1032, Baranowsky, P.
3.
U.S. Nuclear Regulatory Commission, " Collection and Evaluation of Complete and Partial losses of Offsite Power at Nuclear Power Plants,"
NUREG/CR-3992, February 1985.
- 4. U.S. Nuclear Regulator / Commission, " Reliability of Emergency AC Power System at Nuclear Power Plants," NUREG/CR-2989 July 1983.
5.
U.S. Nuclear Regulatory Commission, " Emergency Diesel Generator Operating Experience, 1981-1983," NUREG/CR-4347 December 1985.
6.
U.S. Nuclear Regulatory Commission, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A 44)," NUREG/CR-3226, May 1983.
7.
U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, " Regulatory Guide 1.155 Station Blackout," August 1988.
8.
Nuclear Management and Resources Council, Inc., " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," NUMARC 87 00, November 1987.
- 9. Thadani, A. C., Letter to W. H. Rasin of NUMARC, " Approval of NUMARC Documents on Station Blackout (TAC-40577)," dated October 7, 1988.
- 10. Thadani, A. C., letter to A. Marion of NUMARC, " Publicly-Noticed Meeting December 27, 1989," dated January 3, 1990, (confirming "NUMARC 87-00
- 27, 1989).
Supplemental Questions / Answers." December 32
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- 11. Nuciear Safety Analysis Center, "The Reliability of Emergency Diesel Generators at U.S. Nuclear Power Plants," flSAC-108 Wyckoff, H., ;
September 1986. ,
-12. LaBruria, S., letter to V. S. Nuclear Regulatory Comission Document l
Control Desk, " Station Blackout Coping Analysis," Docket No. 50-354, i dated April 17, 1989.
- 13. Crimins, T. M., Jr., letter to V. S. Regulatory Comission Document Control Desk, " Stat bn Blackout Supplemental Response," Docket No. 50-354, dated April- 30, 1990. i e
- 14. Hope Creek Generating Station Updated Final Safety Analysis Report.
- 15. Crimins, T. M., Jr., letter to U. S. Nuclear Regulatory Comission
- Document Control Desk, " Station Blackout Schedule Comitment," dated June 26,.1990. ,
- 16. Crimins, T. M., Jr., letter to U. S. Nuclear Regulatory Comission ;
Document Control Desk, " Station Blackout Schedule Comitment," dated ,
July 30,-1990.
- 17. Crimins, T. M., Jr., letter to V. S. Nuclear Regulatory Comission .
Document Control Desk, " Station Blackout Revised Schedule " dated March 1, 1991.
- 18. Crimins, T. M., Jr., letter to V. S. Nuclear Regulatory Comission -
Document Control Desk, " Station Blackout Revised Coping Analysis," dated :
March 28, 1991. .
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