ML20213D940
| ML20213D940 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 11/27/1981 |
| From: | Speis T Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0436, CON-WNP-436 NUDOCS 8112180166 | |
| Download: ML20213D940 (53) | |
Text
{{#Wiki_filter:r'. . O ,J N 27 Iggy I ttEMOPMDl!! FOR: Robert L. Tedesco, Assistant Director for Licensing, DL FROM: Themis P. Speis, Assistant Director for Reactor Safety, DSI
SUBJECT:
WNP-2 DRAFT SER g Plant Hame: Docket Number: WNP-2 .[ i, fl' . h .'.'\g g '[S'p[ 7 50-397
, Licensing Stage: OL I e Responsible Branch: Licensing Branch #2 #
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Project Manager: R. Auluck '4 qfh sp./ r _
! DSI Branch Involved: Reactor Systems Branch 4*
Enclosure 2 is the Reactor Systems Branch input to the draft safety evaluation report for WNP-2. This input is based upon a review by Savannah River Plant personnel of the FSAR through Amendment 17 and draft responsts to fomal RSD questions. The reviewers at SRP used the StandFd fieview Plant to conduct the i review, as well as several telecons and a site meeting with the P.SB contract technical monitor and the applicant. RSD sections 5.2.2, 5.4.7. 6.3, and 15 are included as well as the THI-related issues for which RSB has lead responsibility. Please note that fomal documentation of the applicant's responses is required before the final SER can be written. We require this documentation by January 1,1932. The remaining open items are listed in Enclosure 1. In addition several items which we have tentatively resolved in telecons require verification. These are indicated by scoring the right hand margin of the draft SER text. Technical Specifications have also been required in several areas. Although not identified as open areas, these must be verffled as satisfactory during the technical specification review process. n W inar s - >t... Themis P. Speis, Assistant Director for Reactor Safety Division of Systems Integration Distribution e RSB R/ F cc: R. !!attson H. Faulkner Collins R/F D. Eisenhut Section B Members RSB Subject File R. Auluck TSpeis RSB/SLs x WHodges C. P. Ross (Savannah River) jjg, BSheron CONTACT: T. Collins
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. WNP-2 OPEN ITEMS
- l. Overpressurization Protection (5.2.2)- The applicant must submit for our reivew and approval, a plant specific overpressurization analysis using the ODYN code and including the effect of recirculation pump trip.
- 2. Safety / Relief Valve Surveillance (5.2.2)- The applicant must commit to t
participate in a surveillance program to monitor the performance of safety / relief valves.
- 3. Pressure Interlocks on ECC Injection Valves (6.3)- The applicant must verify that interlocks are present at all times for both manual and automatic valve t j operation and that the interlocks do not allow valve opening until the reactor ,
coolant or i rovidepressure an alternative is below the desi n pressure confjguratfon of the ECC which satifies the system requirements involved,SRP of gection6.3
- 4. Premature LPCI Diversion (6.3)- The applicant must rovide assurance that LPCI flow will not be diverted to containment cooling before adequate core cooling is provided. (We have accepted a discussion of emergency procedures and operator training for this item on other applications.)
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- 5. Long Term Air Supply to ADS Valves (6.3)- The applicant must verify that the bottled air supply serving as a backup to the normal air supply :for the ADS valves is valved in during normal operation, or provide justification as to why credit should be given to this air supply, i
- 6. Thermal Power Monitor in Transient Analyses (15)- We require tnat the thermal ,
power nonitor time constant , be included in the plant technical specifications or that no credit be taken for the thermal power monitor in I transient analyses. - l .
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- 7. 00YN Reanalyses (15) - For thermal limit evaluation we require a re-analysis of BWR pressurization transients using the ODYN code.
8, Reclassiffcation of Transients (15)- We require that the turbine trip without bypass and the generator load rejection without bypass events be classified as rnoderate frequency events and they satisfy the MCPR limit of 1.06. s
- 9. Kodification of ADS 1.ogic (II.K.3.18)- We require the applicant to provide one of the following: 1) Analyses.of containtr,ent heatup rates which demonstrate that a high drywell pressure signal will be present at a time early enough to pre:1ude exceeding tne criteria of 10 CFR 50.46 for a stuck open relief valve event or an outside steam line break, 2) a cormitment to nodify the current logic to either bypass the high drywell signal or add a timer to which initiates when Level I water level is reached and which bypasses the high drywell signal upon timing cut. If the timer cption is selected .sn analysis supporting the time setting must be provided.
- 10. Loss of Power to Pump Seal Coolers II.K.3.25- We require verification by the i
applicant of tha applicability of the SWR Owners Group Test Data to WNP-2.
- 11. Restart of Core Spray Systems (II.K.3.21)- We require that modifications be made i to the HPCS system logic so that HPCS will automatically restart on a low water level signal after it has been manually terminated from the control room.
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. .A 5.2.2 Overoressurization Protection The reactor coolant pressure boundary (RCPB) is provided with a pressure relief system to:
(1) Prevent the pressure within the RCPB from rising beyond 110 percent of the design value, and (2) Provide automatic depressurization for small breaks in the nuclear system occurring together with failure of the high pressure' core spray system so that the low pressure coolant injection and the low pressure core spray systems can operate to protect the fuel barrier. The relief system must permit verification of its operability and withstand adverse combinations of loadings and forces resulting from normal, upset, emergency, and faulted conditions. Overpressurization protection at WNP-2 is accomplished through the use of eighteen combination safety relief valves of the Crosby type mounted on the four main steam lines. The following table indicates that the WNP-2 pressure relief system design is similar to other BWR Class 4, 5, and 6 plants. Number of Plant rated pressure steam flow Plant rated Plant Class relief valves (1b/hr) power (MWT) WNP-2 5 18 1.43 x 107 3323
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Grand Gulf 1/2 6 20 ' 1.65 3833 Clinton 1/2 6 16 1.25 2894 # LaSalle 1/2 5 18 1.42 3293 Susquenanna 1/2 4 16 1.35 3293 Fermi 2 4 15 1.42 3293 . Shoreham 4 11 1.05 2436 All of the combination safety relief valves discharge directly to the suppression pool. The valves are designed to meet seismic and quality standards consistent 11/10/81 5-1 Collins /WNP-2 SER 1/A
' , 4 with the requirements of Regulatory Guide (RG) 1.26, " Quality Group Classifica-tions and Standards for Water , Steam , and Radioactive Waste-Containing
- Components of Nuclear Power Plants," and RG 1.29, " Seismic' Gesign Classificatio.5,'
as discussed in Section 3.2 of this report. The basis for overpressure protection in a nuclear reactor is Title 10, C' ode of Federal Regulations, Part 50 (10 CFR Part 50), General Design Criterion (GDC) 15, " Reactor Coolant System Design." This criterion requires that the design conditions of the RCPB are not exceeded during any condition of normal operation, including anticipated operational occurrences. To satisfy this criterion, the overpressurization protection system for WNP-2 was designed in
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compliance with the ASME Pressure Vessel Coda - Section III which requires that the maximum pressure reached during the most severe pressure transient be less than 110 percent of the design pressure. For WNP-2 this pressure limit is 1375 pounds per square inch gauge. The nominal setpoints of the combination safety relief valves are as follows: Setpoint, psig Mode of operation Minimum Maximum Relief (power actuated) 1076 1116 Safety (spring setpoint) 1148 1205 Their total capacity at their set pressure is approximately 112 percent of rated steam flow. Prior to installation, the safety relief valve manufacturer tests the valves hydrostatically for valve response, set pressure, and seat leakage to = certify that design and performance requirements have been met. During the preoperational test program, specified manual and automatic actuation is verified in compliance with Regulatory Guide 1.68, " Initial Test Progr.ams for Water-Cooled Reactor Power Plants." In addition, the applicants have stated the valves will be removed for maintenance and inspection, and tested each , refueling outage in accordance with Section XI of the ASME Boiler and Pressure Vessel Code per plant Technical Specification requirements. 11/10/81 5-2 Collins /WNP-2 SER 1/A
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It is noted that the General Electric Company has agreed to work with the staff and their utility customers to maintain a surveillance program once ,new safety relief valves become operational on any boiling water reactor (NUREG-0152). - Information to be reported will include all abnormalities ranging from minor wear observed during normal inspection to complete failures, including failure to open or close and inadvertent operation. By letter dated from (Washington Public Power Supply System) to (NRC), the applicant has indicated that a safety / relief valve surveillance program is being developed commensurate to that of the BWR Owners Group, the primary objective of which is to gather data to identify generic safety relief valve problems. It is further noted that the applicant is a participant in the BWR Owners Group program to test safety relief valves in compliance with require-ments of Item II.D.1 of NUREG-0737. The applicant has analyzed a series of transients that would be expected to, require pressure relief actuation to prevent overpressurization. These are , tabulated below. Pressurization Events Resultino In Pressure Relief Actuation FSAR Subsection Event 15.1.2 Feedwater Control Failure, Maximum Demand 15.1.3 Pressure Controller, Fail-Open , 15.2.2 Generator Load Rejection, Bypass-On i 15.2.2 Generator Load Rejection, Bypass-Off 15.2.3 Turbine Trip, Bypass-On 15.2.3 Turbine Trip, Bypass-Off ., 15.2.4 Inadvertent MSIV Closure - 15.2.5 Loss of Condenser vacuum
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15.2.6 Loss of Auxiliary Power Transformer - , l 15.2.6 Loss of All Grid Connections 15.2.7 Loss of All Feedwater Flow 15.3.1 Trip of Both Recirculation Pump Motors 15.3.3 Seizure of One Recirculation Pump 4 11/10/81 5-3 Collins /WNP-2 SER 1/A '
The results of these ar.alyses, using t!e nominal valve setpoints, cemonstrate that the maximum vessel pressure will remain below the 1375 psig 1imit. The safety relief valves are direct-acting devices that are assumed to actuate at , the spring setpoint. No credit is assumed for the power operaticn of the valves in the relief mode. For the severe transient of nain steam isclation valve (MSIV) closure with 'a high neutron flux scram, tne maximum vessel bottom pressure is estimated to be 1310 psig when 17 of the 18 safety relief valves are assused to operate in the safety mode. The analysis assured the plant was operating at 105 percent of rated steam flow (14.98 # 108 lb/hr) and a vessel dome pressure of 1020 psig. The analysis was performed using the computer-simulated model described in General Electric Topical Report NEDO-10802, " Analytical Methods of Plant Transient Evaluations for the GE BWR " Comparison of the REDY Code (NEDO-10802) with turbine trip tests at Peach Bottom showed the REDY Code to be nonconserva-tive for overpressurization events. We have reviewed this catter cn a generic basis with the General Electric Company and have evaluated a new calculational basis using the General Electric Company's new computer code 00YN (Ref: Letter from D. G. Eisenhut to holders of CP and OL for BWRs dated January 29, 1981). The applicant has been requested to submit a plant specific overpressurization analysis using tem ODYN for our review and approval. We will report on this analysis in a supplement to this report. Standard Review Plan Section 5.2.2, subsection II.2.c states (for overpressurization analysis) "The reactor scram is initiated either by the high pressure signal or by the second signal from the reactor protection system, whichever is later." The applicant has based the sizing of the safety relief valves on the initia- '- tion of a reactor scram by the high-neutron flux scram which is the second a safety grade scram signal from the reactor protection system following MSIV
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closure. We note, however, that use of the high-neutron flux scram as the second safety grade scram signal has been approved for the Shoreham Nuclear - Power Station and that a modification to Section 5.2.2 of the Standard Review Plan permitting the use of this signal is under review. We believe that the
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qualification and redundancy of reactor protection system equipment coupled with the fact that the reactor vessel pressure is limited to 110 percent of design pressure provides adequate assurance that the reactor vessel integrity 11/10/81 5-4 Collins /WNP-2 SER 1/A
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i will be maintained for the limiting transient event and that the use of trd ,
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second scram signal is acceptable. Accordingly, we find use of the high-neutror. flux scram to be acceptable for WNP-2. The applicant had not included the effects of recirculation pump trip in his . analysis and the initial dome pressure assumed is less than the proposed
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Technical Specification limit. Upon the staff's request, the applicant pro-vided the results of a sensitivity study performed for a BWR/3 to investigate , the effects of increasing the initial reactor pressure relative to the initial value used in the overpressure protection analysis on the peak system pressure. > This analysis showed that increasing the initial operating pressure results in f an increase in peak system pressure which is less than half the initial pressure ; increase. For the WNP-2 project, the proposed technical specification limit ; on the high reactor pressure scram is 1063 psig. Therefore, since the vessel , dome pressure used in the overpressurization analysis was 1020 psig, the maximum [ increase in the initial pressure would be limited to 43 psi, and the maximum peak system pressure increase during the overpressure design transient would be less than 22 psi. Recirculation pump trip has resulted in an increase of 2-6 psi in calculations for other BWRs. These results indicate that adequate ; margin is available to WNP-2 before reaching the code limit and that GDC 15 j will be satisfied even if increased initial dome pressure and recirculation , pump trip are considered. Subject to confirmation by the ODYN reanalyses discussed above, we conclude that the pressure relief system in conjunction with the reactor prcte:tien f system will provide adequate protection against overpressurization of the , j reactor coolant pressure boundary, is in.conformance with the aforeacntioned l t Commission regulations, applicable regulatory guides, and industry standards, ., l and is therefore acceptable. i 5.4.6 Reactor Core Isolation Cooling System i l - The reactor core isolation cooling (RCIC) system is a high pressure reactor coolant makeup system that will operate independently of alternating current
- - power supply. The system provides sufficient water to the reactor vessel to !
I cool the core and to maintain the reactor in a standby condition if the vessel 11/10/81 5-5 Collins /WNP-2 SER 1/A r i I L
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1 a becomes isolated f rom the main concenser and experiences a loss of feedwater flow. The system is also designed to permit complete plant shutdown under cdnditions of loss of normal feed.ater flow by maintaining the necessary , reactor water inventory until the vessel is depressurized to the point where the RHR system can function in the shutdown cooling mode. The RCIC system consists of a steam-driven turnine pump unit and associated valves and piping capable of delivering makeup water to the reactor ves6el I through a head spray nozzle. Fluid removed from the peactor vessel following a shutdown from power operation is normally made up by the feedwater system and supplemented by inleakage from the control rod drive system. If the feeowater system is inoperable, the RCIC system starts automatically when the water level in the reactor vessel reaches the level two (L2) trip setpoint or is started by the operator from the control room. The system is capable of deliverin0 rated flow within 30 seconds of initiation. Primary water supply for the RCIC system come's from the condensate storage tank with a secondary supply from the suppression pool.
; The RCIC system was compared to designs and capacities of similar plant systems l via comparison tables in Section 1.3 of the final safety analysis repert (FSM),
and no unexplained departures from previously reviewed plants were determined. ' RCIC design operating parameters are consistent with expected operational modes as noted in Figure 5.4-10 "RCIC Precess Diagram" of the FSAR, and this complies l with the requirements of GDC 34 regarding residual heat removal. Essential l components of the RCIC system are designated seismic Category I in accordance l with Regulatory Guide 1.29 and Quality Group B in accordance with RegJiatory % i Guide 1.26 as discussed in Section 3.2 of this report. The proposed preopera- = tional and initial test program were reviewed and found to be in conformance l with Regulatory Guide 1.68. ~ l i , The RCIC system is housed within the reactor building which provides protectien ! against wind, tornadoes, floods, and other weather phencmena. Compliance with I the requirements of Criterion 2 of the General Design Criterie in this regard
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l l is discussed in Section 3.8 of this report. Since the condensate storage tank, which is the normal source of water for this system, is not a seismic Category 1 j 11/10/81 5-6 Colli,ns/WNP-2 SER 1/A l __ _ . _ . _ ,--. ____m. _ _ , . _ . - -
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5 t g structure, the applicant has committed to provide an automatic safety grace suction switchover to the suppression pool to ensure a water supply in tne event of a safe shutdown earthquake and concomitant failure of the condensate storage tank. In addition, the system is protected against pipe whip inside and outside containment as required by Criterion 4 of the General Design Criteria, as discussed in Section 3.6 of this report.
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The high pressure core spray and reactor core isolation cooling systems are located in different rooms of the reactor building for additional protection against common mode failures. They use different energy sources for pump motivation (steam turbine for reactor core isolation cooling pump, electric power for high pressure core spray pump) and different power systems for control power. This diversity conforms to the requirements of Section 5.4.6 of the Standard Review Plan. To protect the RCIC pump from overheating, the reactor core isolation cooling system contains a miniflow line which discharges into thd suppression pool when the line to the reactor vessel is isolated. When sufficient flow to the vessel - is achieved, a valve in the miniflow line automatically closes, thus directing all flow to the reactor. The reactor core isolation cooling system is protected against the effects of water hammer when starting by a jockey pump system which maintains the discharge piping filled up to the injection valve. A high point vent in provided and the system will be checked at least once every 31 days to l assure that the lines are filled. The reactor core isolation cooling system includes a full flow test line with water. return to the condensate storage tank :3 for periodic testing. Technical specifications include a flow test at least every 92 days and a system functional test at least every 18 months with 1 sinulated automatic actuation and verification of proper automatic valve ~ position. In both tests, verification is obtained that the reactor core isolation cooling pump will dtvelop a minimum flow of 600 gallons per minute. Isolation between the reactor coolant system and the reactor core isolation - cooling system is previded by: (1) two check valves and a closed de powered valve in the reactor core isolation cooling system discharge line, and (2) two 11/10/81 5-7 Collins /WNP-2 SER 1/A L.
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normally open motor-operated valves in the steam line to the reactor cort isolation cooling steam turbine. We require that the motor operated vaives be classified Category A and the check valves Category A/C in secordance witn tne provisions of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and that these valves be leak testad' periodically. Specific testing requirements for these valves are discussed in Section 2.
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The reactor core isolation cooling system has controis which can shut down the system if operating conditions exceed certain limits. We requested the appli-cant to describe the provisions that will prevent the reactor core isolation cooling system from being inadvertently isolated or shutdown due to spurious signals. The applicant stated that the differential temperature trip setpoint is based upon the normal operating temperatures plus an allowance for heat from a predetermined steam leak. Maximum temperature swings from local meteorological data have also been taken into account. The differential setting will be adjusted if required during startup testing. We find this acceptable. Sperious isola-tions due to high initial steam flow in the turbine steam lines is discussed under II.K.3.15 of NUREG-0737 in Chapter . The applicant was also requested to identify the portions of the reactor core isolation cooling system that have a low design pressure and therefore requi.re either relief devices to comply with Section III, Article NB-7000 of the ASME Boiler and Pressure Vessel' Code or a basis for deciding that relief devices are not required to protect these lines from overpressurization during normal - plant operation. The low pressure lines identified by the applicant, in addition to the normal isolation valves, have either relief valves, rupture ciscs or are vented to provide backup protection. We find this acceptable. .= The reactor core isolation cooling system is capable of supplying coolant to the reactor following feedwater isolation and reactor shutdown under n.ormal - and accident conditions. The reactor core isolation cooling system conforms to the requirements of General Design Criterion 5 in that the_RCIC system is _
"10 be provided by LPM.
11/10/81 5-8 Collins /WNP-2 SER 1/A
. e not shared between units; Criterion 29 (in conjunction with the HPCS System' j through quality controlled construction and periodic testing; Criterion 33 (in conjunction with HPCS) in that operation with only offsite or only onsite powet is possible for protection against small breaks; Criterion 34 (in conjunction with HPCS) in that residual heat removal while still at high pressure can be accomplished assuming a single' failure and with or without offsite power; and Criterion 54 in that suitable leak detection and isolation capability is pro-vided on piping penetrating containment. Review of the drawings, component descriptions, and design criteria for the reactor cord isolation cooling system were conducted and, on the basis of this review, we conclude that the design of the reactor core isolation cooling systems conforms to the Commission's regulations and to the applicable regulatory guides, and is therefore acceptable.
5.4.7 Residual Heat Removal System The residual heat removal (RHR) system comprises three independent loops; each loop contains its own motor-driven pump, piping, valves, instrumentation, and controls. Loops A, B, and C each have a suction source from the suppression pool and each is capable of discharging water to the reactor vessel via a separate nozzle or back to the suppression pool via a full flow test line. In addition, the A and the B loops have heat exchangers which are cooled by the standby service water system. They can also take suction from the reactor recircula-tion system, and can discharge into the reactor recirculation discharge line, or to the suppression pool, or to the drywell spray spargers. Loops A and B also have connections to reactor steam via the reactor core isolation cooling (RCIC) steam line and can discharge condensate to the RCIC pump suction or to the suppression pool. . m The RHR system is used in conjunction with the main steam and feedwater systems (main condenser), or with the RCIC system in conjunction with the safety / relief ~ valves to cool down the reactor coolant system following shutdown. In the event - the RHR isolation valves are inoperative, an alternative shutdown method is used. In this method, water is pumped from the suppression pool through the RHR heat exchangers and into the reactor vessel. The vessel water is allowed ~ to overflow the steam lines and discharge back to the suppression pool via the 11/10/81 5-9 Collins /WNP-2 SER 1/A
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discharge lines of those safety relief valves which are part of Automatic Depressurization System. The RHR system operates in five different modes: (1) Shutdown cooling, (2) Steam condensing, t (3) Suppression pool cooling, (4) Containment spray cooling, and (5) Low pressure coolant injection. All five modes of operation use the same hardware. Only the shutdown cooling and the steam condensing modes are covered in this section. Modes (3), (4), and (5) are reviewed through other sections of the Standard Review Plan (6.2, 6.3). The normal operational mode of the residual heat removal system is the shutdown cooling mode, which is used to remove decay heat from the reactor core to achieve and maintain a cold shutdown condition. The steam condensing mode is used to condense steam while the reactor is isolated from the main condenser and vessel level is being maintained by the reactor core isolation cooling system. The heat removed in the RHR heat exchangers is transported to the ultimate heat sink by the RHR service water system. - The RHR system was compared to designs and capacities of such systems in similar plants and no unexplained departures from previously reviewed plants were found. - The RHR system is designed to operate, with or without offsite power with a - single active failure. Control of the RHR system is accomplished from the control room. Using the system process diagrams, P& ids, system safety analysis, and component performance specifications, it was determined that the system provided at WNP-2 has the capacity to bring the reactor to cold shutdown 11/10/81 5-10 Collins /WNP-2 SER 1/A
conditions in a re:sonable period of time [ assuming operation of only , safety grade equipment. - Isolation between the RHN suction line and the reactor coolant system recirculation loop is provided by an inside containment isolation valve and an outside containment isolation valve. Each valve is interlocked with a separate switch which prohibits opening of the associated valve if the recirculation loop pressure exceeds the shutdown range. These same interlocks initiate valve closure on increasing reactor pressure. An operator error cannot open either valve at a pressure above the shutdown range. Isolation between the RHR shutdown return to the recirculatien discharge line and the RHR connection to the RCIC head spray both have a check valve and globe valve at the pressure boundary. The globe valve has a pressure interlock which prevents opening of the valve due to operator error when the higher pressure system exceeds the shutdown range. This same interlock initiates valve closure on increasing reactor pressure. We require that the motor-operated valves be classified Category A and the check valves Category A/C. in accordance with Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and that they be tested periodically. Specific testing requirements for these valves are discussed in Relief valves are provided in each of the low-
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Section 3.
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pressure lines that interface with the reactor coolant system to protect against overpressurization from leakage from the reactor coolant system. Isolation of the reactor coolant system from the shutdown cooling mode portions of the resid-ual heat removal system in this manner complies with the requirements of Section 5.4.7 of the Standard Review Plan and is acceptable. The low pressure et coolant injection mode pressure isolation provisions are discussed in Section 6.3. . _ The drywell and suppression pool spray lines have no pressure isolation require-ments from the reactor coolant system. The containment isolation requirements of the residual heat removal system are discussed in Section 6.2 of this report. -
"To be supplied by LPM.
11/10/81 5-11 Collins /WNP-2 SER 1/A
The residual heat removal system is designed to the seismic Category I require-ments of Regulatory Guide 1.29 as discussed in Section 3.2 of this report. It is housed in the reactor building for protection against the effects of floooing, tornadoes, hurricanes, and other natural phenomena. Compliance with Criterion 2 of the General Design Criteria in this regard is discussed in Section 3.8 of this report. Compliance with Regulatory Guide 1.26 regarding Quality Group classifications is discussed in Section 3.2 of this report. The containment isolation requirements of Criteria 55, 56, and 57 of the General Design Criteria are discussed in Section 6.2 of this report. Systems used for cooling the residual heat removal system conform to the requirements of - Criteria 44, 45, and 46 of General Design Criteria, as discussed in Section 9.2 of this report. Those portions of the residual heat removal system which are also part of the emergency core cooling system are designed to operate under both normal and accident conditions. The system is protected against missiles (discussed in Section 3.
- of this report) cnd pipe whip (discussed in Section 3.6 of this report). In this way, the residual heat removal system complies with the requirements of Criterion 4 of the General
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Design Criteria. There is only a single line from the recirculation system to the residual heat removal system for use in cooling the reactor in the shutdown mode. This line is vulnerable to a single failure of either of the isolation valves. The appli-
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cant has an alternate cooling path using the safety relief valves and suppression pool cooling in the event of a failure in the suction line which would preclude residual heat removal system operation. Both paths are provided with emergency power supplies. To assure the long-term operability of the automatic depres-surization system (ADS) valves for the alternate shutdown cooling mode, 2' compressed air cylinders are provided as part of the containment instrument :=- air system. The associated valves and all the interconnecting air system piping from the isolation valve outside containment to the ADS valves are designed to the requirements of ASME Section III, Class 2 and 3, as applicable, --
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and are seismic Category 1. These alternate cooling provisions satisfy the single failure requirements of Criterion 34 of the General Design Criteria. -
*To be supplied by LPM.
11/10/81 5-12 Collins /WNP-2 SER 1/A
r, 6.3 Eme'roency Core Cooling System The emergency core cooling system (ECCS) is designed to provide water to the reactor coolant system in the event of'a brea'k in the pressure boundary. The ECCS capability extends to failures as large as a double-ended rupture of the largest pipe carrying water or steam, and spurious safety / relief valve operation. The basis for the design of the ECCS is tol11mit damage to the fuel cladding in accordance with Title 10, Code of Federal Regulations, Part 50.46 (10 CFR 50.46). The system must be capable of performing its design fun (tion without offsite power and with a single failure, including loss of an emergency diesel. 6.3.1 System Description The ECCS consists of the following systems: (1) High Pressure Core Spray System (HPCS) (2) Automatic Depressurization System (ADS) (3) Low Pressure Core Spray System (LPCS) (4) Low Pressure Coolant Injection System (LPCI) 1 ~ The HPCS is provided to maintain the reactor vessel water level above the tcp
- of the active core in the event. of pipe breaks of one inch diameter or smaller and to provide spray cooling in case the core is uncovered. Activation of the HPCS does not require the depressurization of the reactor vessel. The system includes a single motor-driven centrifugal pump which takes suction from the condensate storage tank or the primary containment suppression pool. An automatic switching feature is provided. HPCS flow is dependent on the pressure differential that exists between the reactor system and the suction source.
l Pump characteristic curves indicate a flow rate of approximately 2500 gpm against J-
.-. -- the safety valve setpoint pressure (1148 psi) and that the rated HPCS flow; ,
(6350 gpm) is attained at approximately 360 psid. HPCS discharges water into the ' '
- reactor via a spray sparger mounted on the reactor vessel internal wall above the
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G core. The HPCS system is designed to operate from normal offsite auxiliary AC , power or from its own diesel generator.
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The ADS is provided to depressurize the reactor coo'lant system in the event a small pipe break occurs and the HPCS system cannot maintain reactor vessel water level or fails to start. J The ADS employs 7 of the 18 safety / relief valves to reduce system pressure so that the low pressure systems may inject water to cool the core.
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The LPCS system is provided to replace reactor vessel water inventory and to supply spray cooling following large pipe breaks in which the core may uncover. The system includes a motor-driven centrifugal pump which takes suction from the suppression pool and discharges water to the reactor vessel via a spray sparger mounted on the reactor vessel internal wall above the core. Pump characteristic curves indicate that rated LPCS flow (6350 g::m) is at+=ined at a pressure of approximately 260 psid. and that the pump cut off head is approximately 350 psid. The LPCS sparger is separate from the HPCS sparger. The LPCS system is desicned Ic. operate from nonnal auxiliary AC power or from the standby AC power system. The LPCS pump and RHR pump A are on the same electrical bus and are sup-plied emergency power by the same diesel generator. The LPCI system is provided to replace reactor vessel water inventory following large pipe breaks. The system is an operating mode of the Residual Paat Removal (RHR) system which consists of three independent loops (A, B, and C). - Each loop has a motor-driven pump (7450 gpm) which takes suction from the suppression pool and supplies water to the reactor vessel via a separate nozzle through the reactor vessel wall. In addition, loops A and B can also take suction from the reactor recirculation system suction or fuel pool, and can discharge into the reactor via a feedwater line, fuel pool cooling discharge, or to the containment spray spargers. RHR loops A and B have heat exchangers which are cooled by the RHR service water system and are used to transfer the decay heat from the reactor core to the ultimate heat sink. The three LPCI (RHR) pumps are powered from AC power buses having standby power source backup supplies. RHR pumps B and C are on the same electrical bus and receive 6-2 -
emergency power from the same diesel generator. RHR pump A is on the same electrical bus as.the LPCS pump. 6.3.2 Evaluation 6.3.2.1 Single Failures We reviewed the system description and piping and instrumentation drawings to assure that abundant core cooling will be provided during the injection phase with and without offsite power and assuming a single failure as required by Criterion 35 of the General Design Criteria. A low reactor vessel water level and/or high containment pressure signal is required to start pumps and open discharge valves. The applicant provided in Section 6.3.3 of the Final Safety Analysis Report an analysis to demonstrate that the most lir t .1g break size, break location, and single failure had been considered for WiiP-2 The most limiting combinations are tabulated below.
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Break Size Break Location Single Failure Systems Remaining Small and Recirculation Suction HPCS ADS, LPCS and all Intermediate Line LPCI Recirculation Suction LPCI Diesel. HPCS., ADS, LPCS and Large Line Generator one LPCI The applicant has analyzed main steam breaks inside and outside containment, HPCS line breaks and feedwater line break ' locations. The analyses have shown these break locations, assuming the worst single failure, are not limiting. These breaks all occur at higher elevations which result in faster depres-surization and earlier actuation of the emergency core cooling system. In addition, the reduced loss of inventory results in lower peak clad temperature. 6-3
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6.3.2.2 Oualification of the Emergency Core Cooling System The emergency core cooling system is designed to meet seismic Category I requirements in compliance with Regulatory Gu'ide 1.'29, " Seismic Design Classi-fication," as discussed in Section 3.2 of this report. It is housed in structures designed for seismic events, tornadoes, floods, and other phenomena in accordance with the requirements of Criterion 2 of the General Design Criteria as discussed in Section 3.8 of this report. Emergency core cooling system equipment is designed in compliance with Regulatory Guide 1.26, " Quality Group Classification and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power P1 ants," as discussed in Section 3.2 of this report. Protection of the emergency core cooling system against pipe whip and against discharging fluids in compliance with the requirements of Criterion 4 of the General Design Criteria and Regulatory Guide 1.46, " Protection Against Pipe Whip Inside Containment," is discussed in Section 3.6 of this report. Evalua-tion of the instrumentation and controls for the emergency. core cooling system is discussed in Section 7.3 of the report. Compliance with the inservice inspection requirements of Criterion 36 of the General Design Criteria is discussed in Section 6.6 of this report. Environmental qualification of the emergency core cooling system equipment for operation under normal and accident conditions as required by Criterion 4 of the General Design Criterion is discussed in Section 3.11 of this report. l
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1 6.3.2.3 Functional Design The available net positive suction head for the pumps in the emergency core cooling system has adequate margin to prevent cavitation and assure pump operability in accordance with Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps." We requested that the applicant also provide calculjtions verifying that flashing would not occur at any point in the ECCS suction lines as a re-sult of local elevation changes in the piping runs. The applicant provided supplementary calculations which assumed conservative suppression pool tempera-ture and friction losses. The results indicate that at all points, the pressure calculated exceeded the vapor pressure of the fluid. This is acceptable to us. The HPCS incorporates relief valves to protect the components and piping from inadvertant overpressure conditions resulting from either thermal expansion or backpressure leakage into the low pressure portions of the system. The LPCS and LPCI systems are not designed to withstand normal reactor operating pressure. Each of the low pressure lines that interface with the reactor coolant system. has a testable chceck valve inside primary containment backed up by a normally closed motor-operated gate valve cutside of containment. Relief valves are provided in the low pressure lines to protect against leakage from the reactor coolant system. An interlock is provided on the motor operated valves which prevents their opening until the differential pressure across the valve is below a specified value. We require that this interlock be present at all times for both automatic and manual valve actuation, and that the setpoints be such that the valve cannot be opened until reactor coolant pressure is below that of the low pressure ECC system involved. We will report on this matter in a supplement - on this report. Containment. isolation in accordance with the requirements of l Criterion 55 of General Design Criteria is discussed in Section 6.2 of this report. Th l
periodic testing and leak rate criteria for these valves that isolate the reactor coolant system from the emergency core cooling system are discussed in Section 3.9.6 of this report. The detection of leaks from those portions of the emergency core cooling system within primary containment is discussed in Sections 5.2.5 and 9.3.3. For portions ,of the emergency core cooling system outside of primary containment leak detection monitfrs are provided in each ! of the water-tight equipment rooms to detect leakage and alert the operator to possible flooding conditions. All the emergency core cooling systems have miniflow lines to permit a limited amount of ficw in the event an isolation valve between the reactor coolant system and emergency core cooling system is closed, for any reason, in order to protect the pumps from overheating. When sufficient flow passes through the injection lines, valves in the miniflow lines automatically shut, divert- - ing all flow to the pressure vessel. The lines from the suppression pool to the suctions of the low pressure coolant injection and low pressure core spray j pumps each have an open motor-operated valve outside of containment with controls arranged so that a key is required to unlock a lever to close the valve. The suction of the high pressure core spray from the suppression pool contains a closed motor-operated valve outside the containment designed'to open so that the system automatically pumps water trom the suppression pool instead of the condensate storage tank when the condensate storage tank water is exhausted. As a backup to tha high pressure core spray system, the automatic depressuriza-tion system can be used to depressurize the system and allow the functio.ning of the low pressure cooling systems in the event of a small break. The air supply to the automatic depressurization system valves is provided in accident conditions by seismically qualified accumulators backed up by a long-term bottled air supply to compensate for leakage past accumulator check valves. Nitrogen bottles located outside of containment can be readily changed ont to extend the air supply indefinately.We require that this air supply be connected at all times during normal operation. ~
One of the design requirements of the emergency core cooling system is that cooling water flow be providea rapidly fowlloing the initiation signal. By' always keeping the emergency core cooling system pump discharge lines full, the lag time between the signal for pump start and the ini+.iation of flow into the reactor pressure vessel can be minimized. In addition, full discharge lines' will prevent potentially damaging water hamer occurrences on system startup. At WNP-2 a fill system consisting of a jockey pump in the RCIC system and in each
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of the ECCS subsystems (except ADS) is provided. Maintenance of the filled y a status of the system is ensured by continuous indication of oumo operation and a pump discharop nrou nra. In accordance with monthly surveillance procedures the vent lines in the filled systems are opened and checked for flow to eliminate the possibility of the fomation of air pockets. Pressure instrumentation M provided on the jockey pump discharge line initiates an alarm in the main F 4 control room when pressure in the discharoe line is less than the hydrostatic j head required to maintain the line full of water up to the in.iection valves, h
,1 The emergency core cooling system pumps must have the capability to operate for an extended period of time during the long-term recirculating cooling phase following a loss-of-coolant accident. The applicant has provided pump reliability information based upon actual operation experience on similar pumps manufactured by Ingersall-Rand, the WNP-2 supplier. As discussed in Section 5.4.7 we are awaiting additional information from the applicant on deep draft pump reliability.
i Safety / relief valve operability will be demonstrated during toe power ascension i phase of the plant startup test program by manually actuating each safety / relief valve (including the ADS valves) one at a time to measure discharge capacity and ! to demonstrate that no blockage exists in the valve discharge line. After commercial turnover all of the safety relief valves will be tested in accordance with Section XI, Article IWV of the ASME Boiler and Pressure Vessel Code. The applicant has also stated that direct valve position indication, via acoustic
monitors in the tail pipe section, will be provided. This is discussed under item II.D.3 in Chapter 22. The staff asked the applicant to provide assurance that the safety / relief valves have been qualified by environmental testing to support the assumption that 6 of the 7 ADS valves will operate. This is discussed in section 3.11 of this t report. 6.3.3 Testina The applicant states that operability of the emergency core cooling systems will be demonstrated by preoperational and periodic testing as required by Regulatory Guide 1.58, " Initial Test Programs for Water-Cooled Reactor Power Plants," and Criterion 37 of the General Design Criteria. 6.3.3.1 Preeperational Tests Preoperational tests will assure proper functioning of controls, instrumenta-tion, pumps, piping, and valves. Pressure differentials and flow rates will be measured for later use in determining acceptable performance in periodic tests. The applicant has committed to meet the guidelines of Regulatory Guide 1.68 mentioned above for preoperational and initial startup testing of the emergency core cooling system as noted in Section 14 of this report. 6.3.3.2 Periodic Comoonent Tests We will ' require the applicant to test the subsystems comprising the emergency core cooling system (except for the automatic depressurization system) every 92 days to show that specified flow rates are attained. Also, we will require every 18 months that a test be performed in which all subsystems are actuated through the emergency operating sequence. These tests comply with Criterion 37 of the. General Design Criteriac
l e 6.3.4 Performance Evaluation We reviewed the loss-of-coolant accident analyses presented by the applicant in Section 6.3.3 of the Final Safety Analysis' Report. Calculations were conducted in accordance with the methods described in General Electric Topical Report NEDO-20566, " General Electric Company Analytical Model for Loss-of- . Coolant Analyses in Accordance with 10 CFR Part 50, Appendix K," dated August 1974 and " General Electric Refill Reflood Calculation" transmitted December 20, 1974. During 1977, the General Electric' Company propoged several changes to its emergency core cooling system evaluation model. These changes have been approved by us and are described in our report " Safety Evaluation for General Electric Emergency Core Cooling System Evaluation Model Modifications." -These methods constitute an evaluation model that conforms to the requirements of Appendix K to 10 CFR Part 50.
, The results of the break analyses along with the results of the review of each
'CCS system were used to verify that the proposeo ECCS complied with the five t
major acceptance criteria outlined below. Compliance with the first three criteria are demonstrated analytically. Compliance with the coolable geouletry ano long-term cooling criteria are demonstrated more intuitively. Coolable geometry is maintained if the first two criteria are met. Long-term cooling capability is verified by the
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composite review of the ECCS systems and the various support systems, i The five major acceptance criteria for the emergency core cooling system as speci-fied in 10 CFR 50.46 are: . (1) The calculated maximum peak cladding temperature shall not exceed 2200 F. (2) The calculated total oxidation of the cladding shall nowhere exceed i 0.17 times the total cladding thickness before oxidation. (3) The calculated. total amount of hydrogen generated from the chemical reaction of the clad. ding with wat.er or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in 6-9
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the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. (4) Calculated changes in core geometry shall be such that the core remains amenable to cooling. - - (5) The calculated core temperature shall be maintained at an acceptable low value and decay heat shall be removed for the extenged period of time required by the long-lived radioactivity remaining in the core after any calculated successful initial operation of the ECCS. The applicant has demonstrated compliance with the first three of these' criter'ia as follows: Maximum values From Break Analyses Allowable Peak Cladding Temperature (PCT) 1960 2200'F Maximum Cladding Oxidation 1.1% 17% Maximum Total Hydrogen Generation .07% 1% A coolable geometry is demonstrated by the compliance with the criteria for l the PCT and the maximum cladding oxidation as discussed in NEDO-20566. l Long-term cooling is assured by the use of redundant systems which have adequate water sources available to remove the decay heat generated within the reactor core and transfer the heat to the ultimate heat sink. No single failure was identified that would prevent the ECCS from meeting this criterion. The systems are designed to reflood the reactor core to at least the 2/3 core l level and maintain this level even under the most adverse circumstances. The major equipment for each system; other than the ADS, is located in separate water-tight rooms outside primary co.ntainment. O
9 Results of the break spectrum analyses have shown that the limiting break is a design basis accident in a recirculation suction pipe. Coincident with the break, failure of the Division 2 emergency diesel generator (LPCI loops B and C are powered by this diesel generator) is assumed to occur. All other breaks resulted in lower peak cladding temperatures. The applicant has indicated that some operator acti6ns are assumed in the loss-of-coolant accident analyses 10 minutes after accident initiation. Section 6.3 of the Standard Review Plant states that no credit for operator actions should be taken prior to 20 minutes. The actions assumed are initiation of suppression pool cooling during the design basis accident and manual depressuriza-
' tion of the reactor coolant system for a main steamline break outside of contain-i ment. The applicant has referenced an analysis performed on a similar BWR/5 plant (La Salle) for limiting outside steamline break where operator action is taken at 20 minutes. The results indicate a peak cladding temperature of 1250*F. We find this acceptable based upon the similarity of plant design and the wide margin before reaching the limit of 2200 F. The startup of suppression pool cooling 10 l minutes into the event has been accepted by us on previous applications. WNP-2 l
1 emergency procedures instruct operators to start suppression cooling when pool temperature exceeds degrees Farenheit. This condition is expected very shortly after a loss-of-coolant accident and would be anticipated by the operator. We con-clude on the basis of the simplicity of the action assumed and the emergency pro-cedures on containment control that the operator action assumed in the loss-of-coolant accident analysis for WNP-2 are acceptable. O
s The applicant also discussed the simultaneous closure of a recirculation flow control valve during a loss-of-coolant accident. The applicant's basis for the closure time was an electronic velocity limiter designed to limit the open-ing and closing rate to 11 percent per second. The valve controller is not V classified as equipment which is essential to safety. Therefore, the controller is not scrutinized by us to the same extent that a component required for safety b would be. However, for the loss-of-coolant accident, the valve is not called A upon to actively mitigate the consequences of the accident, but is only needed to I passively remain in its current position. If the control valve were in the F automatic mode at the time of the loss-of-coolant accident, the control system y would nonnally call for the valve to open. If the control valve were in the j' manual mode, operator action would nonnally be required to close the valve. Two ( failures in the control logic would be required for the valve to close even at the 11 percent per second rate. Further, the control logic is outside the drywell / and is not subject to the loss-of-coolant accident environment. Therefore, the 11 percent per second closure rate, which will be verified by periodic tests i required by plant technical specifications, is a reasonable limit on valve closure ! rate for the loss-of-coolant accident analysis. The applicant indicated that if this additional failure is included in the loss-of-coolant accident, the peak f temperature for the worst break would be increased approximately 50 degrees Fahren-heit which still satisfies the criterion of 10 CFR Section 50.46. Therefore, we find the loss-of-coolant accident analyses to be acceptable. l
The low pressure coolant injection flow may be diverted manually to drywell . L spray cooling cr to suppression pool cooling. The applicants i F have stated that the WNP=2 emergency procedures contain adequate y' cautions to deter the operator from premature' flow diversion. These proce-dures, which are based on guidelines accepted by us (see Chapter 22 Item I.C.1), caution the operator against diversion unless adequate core cooling is assured. LPCI diversion is identified in the procedure as secondary to core cooling requirements except in those instances outside the design envelope involving multiple failures for which maintenance of containment integrity is required to minimize risk to the environment. We have, reviewed the containment res,ponse analyses for the design basis event to determine the need for low pressure coolant injection diversion. These analyses indicate that there should be no need for wetwell spray actuation in the time frame during which the peak cladding temperature is reached. The operator's focus would, therefore, be on maintaining core cooling. Based on these analyses and the emergency proceduces discussed above, we find the applicants' position on low pressure coolant injection diversion to be acceptable. Review of all emergency procedu.res is b'eing addressed in Items I.C.1'and I.C.B of Chapter 22 of this report. The core spray sparger for both the high and low pressure core spray systems each consists of two semicircular segments which form an essentially complete circular sparger. Water is sprayed radially onto the tops of the fuel assem-blies by short elbow nozzles spaced around the sparger. Tests of this type of spray system were performed in a full-scale test in which air at atmospheric pressure simulated the post less-of-coolant accident steam environment and indicated adequate cooling was delivered to each fuel assembly. However, recent tests conducted on a single nozzle indicate that the actual steam environment may adversely affect the distribution of flow from certain types of core spray nozzles. As discussed in NUREG-0410, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," this problem is beit.g studied by us under Task Action A-16 entitled, " Steam Effects on BWR Core Spray Distribution." Prefiminary analyses .and measurements have been made which support the existence of a significant safety margin between that ', 6-11
p. amount of spray flow provided to each fuel assembly in the post loss-of-coolant accident steam environment and that used to calculate the spray cooling , coefficients assumed in the loss-of-coolant accident analyses. Tests have racently been conducted by General Electric to confirm spray flow margins used in the emergency core cooling system los's-of-coolant accident analyses. We will review the results of these tests when they are submitted. In the interim period, we believe there is a sufficient technical basis to permit licensing of WNP-2. This conclusion is based on: (1) The existence of safety margins between available and required spray flow indicated by preliminary analyses and neasurements, (2) The existence of counter-current-flow-limiting phenomena should provide a steam / water layer on top of the core anc force even distribution of the core spray, (3) The timely confirmation of the spray flow margin presently believed to exist which should be provided by the aforementioned tests and information. 6.3.5 Conclusions We reviewed piping and instrumentation drawings and the description of the emergency core cooling system presented in the Final Safety Analysis Report. We find the design of the system acceptable because it conforms to the pertinent Regulatory Guides, Standard Review Plan and General Design Criteria. In addition, based on the discussion above, we find the performance of the emer-gency core cooling system acceptable because it conforms with the requirements of 10 CFR Section 50.46 pending resolution of the issue noted above.
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15 SAFETY ANALYSIS Introduction , Two groups of design basis events are evaluated in this section: anticipated operational occurrences and accidents. In order for $he analysis of events in either group to be acceptable, it is required that a conservative model of the reactor be used, and that all appropriate systems whose operation (or postulated misoperation) would affect the event be included. Anticipated operational occurr.ences are expected to occur during the life of the power plant and are analyzed to assure that they will not cause damage to either the fuel or to the reactor coolant pressure boundary and to assure that the radiological dose is maintained within Title 10,' Code of Federal Regulations, Part 20 (l'0 CFR 20) guidelines. Design basis accidents are not expected to occur, but are postu-lated because their. consequences would include the potential for the release of significant amounts of radioactive material. They are analyzed to determine the extent of fuel damage expected and to assure that reactor coolant pressure boundary damage, beyond that assumed initially by the design basis accident, will not occur, and that the radiological dose is maintained within 10 CFR 100 guidelines. This is done for the boiling water reactor control red drop accident by requiring that the calculated peak fuel enthalpy remain below 280 calories per gram, the limit used in Regulatory Guide 1.77, " Assumption Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," and accepted by us for use as a fuel safety limit for boiling water reactors. The 280 calories per gram energy density value will provide a conservative maximum limit to ensure that core damage from this postulated event will be acceptable and that both short term and long term core cooling capability will not be impaired.
. . a For loss-of-coolant accidents, the acceptance criteria for the emergency core cooling system specified in Title 10, Code of Federal Regulations, Part 50.46 15-1
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are: (1) the peak cladding temperature must remain below 2200 degrees Fahrenheit; (2) maximum cladding oxidation mast nowhere exceed 17 percent of the total cladding thickness before oxication; (3) total hydrogen generation must not exceed one percent of the hypothetical amount that would be generated if all the metal in the cladding cylinders, excluding the cladding surrounding the plenum volume, were to react; (4) the core must be maintained in a coolable geometry; (5) calculated core temperatures after successful initial operation of the emergency core cooling system shall be maintained acceptably low and decay heat shall be removed for the extended period o& time required by the long-lived radioactivity remaining in the core. 15.1 Abnormal Ooerational Occurrences Abnormal operational occurrences are those transients resulting from single equipment failures or single operator errors that might be expected to occur during' normal or planned modes of plant operation. The acceptance criteria j for these transients are based on General Design Criteria 10 and 15. i Criterion 10 specifies that the reactor core and associated control and instrumentation systems be designed with appropriate margin to assure that - l acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. Criterion 15 specifies that sufficient margin shall be included to assure that design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences. Specific acceptance criteria (Standard Review Plan) for the moderate frequency transients are: (1) Pressures in the reactor coolant and main steam systems should be maintained below 110% of the design values according to ASME Boiler and Pressure Vessel Code, Section III, Article NS-7000, " Overpressure Protec-tion." For WNP-2 which has a design pressure of 1250 pounds per. - square inch gauge, the pressure should not exceed 1375 pounds per square . inch gauge during any anticipated operational occurrence. Grand Gulf SER 15-2 T. collins /Siegel E
(2) Fuel cladding integrity should be maintained by ensuring that the reacto,r core is designed with appropriate margin during any conditions of normal operation, including the effects of anticipated operational occurrences. For boiling water reactors, the minimum value of the critical power ratio reached during the transient should be s0ch that 99.9% of the fuel rods in the core would not be expected to experience boiling transition during core wide transients. This limiting value of the minimum critical power ratio, called the safety limit, will vary for different plants and/or product lines. For WNP-2 the value is 1.04 (see Section 4.4). For transients that affect only a portion of the core, the same value of the safety limit is used to provide additional conservatism. . (3) An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently. (4) An incident of moderate frequency in combination with any single active component failure, or operator error, should not result in loss of function of any barrier other than the fuel cladding. A limited number of fuel rod cladding performations is acceptable (see II.K.3, Item 44 of Chapter 22) The applicant used the following conservative assumptions with respect to initial power, scram reactivity, reactivity coefficients, and power profiles in the analyses: (1) Initial power 104.4 percent of rated power (corresponding steam flow rate 105 percent of rated steam flow rate). (2) Scram reactivity characteristics accounting for end-of-cycle conditions l which result in the most conservative effects. The slowest allowable control rod scram motion is assumed with a scram worth shape for all the control rods fully withdrawn at the end-of-cycle. ( (3) Scram reactivity calculations incorporating a 20 percent safety ; ! conservatism factor (i.e. , rod worth of 0.80 x end-of-cycle rod worth). l I l 15-3 i l
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This includes the effect of a stuck control rod as required by General Design Criterion 26. (4) Core burnup selected to yield the most limiting expected combination of moderator temperature coefficient, void coefficient, Doppler coefficient, axial power profile, and radial power distributien. The transients analyzed involved the following reactor scrams which satisfy the requirements of Criterion 20 of the General Design Crv(eria: (1) Reactor vessel high pressure, . (2) Reactor vessel low water level, (3) Turbine stop valve closure, (4) Turbine control valve fast closure, (5) Main steam line isolation valve closure, , i (6) Neutron monitoring system scram. , Time delays to trip for each scram signal were included in the analyses. The transient events were categorized in terms of the following system parameter variations: -
- 1. Decrease in Core Coolant Temperature l
Transients analyzed in this group included loss of feedwater heaters, feedwater control failure to maximum demand, pressure regulator failure in the open direction, inadvertent opening of a safety / relief valve and
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inadvertent residual heat removal shutdown cooling operation. , e . e 15-4 4 mm me mo me ee e .
- 2. Increase in Reactor Pressure '
Transients in this group included pressure controller failure-closed-generator load rejection and turbine ttip with and without turbine bypass, inadvertent MSIV closure, loss of condenser vacuum, loss of auxillary power transforcer, loss of all grid connections, loss of all feedwater flow, and failure of residual heat removal shutdown cooling.
- 3. Decrease in Reactor Coolant Svstem Flow Rate /
Transients in this group included trip of one or both recirculation pump motors and recirculation flow control failure to decrease flow.
- 4. Reactivity and Power Distribution Anomalies Transients in this group included rod withdrawal error at power, abnormal startup of an idle recirculation loop and recirculation flow controller failure with increasing flow.
- 5. Increase in Reactor Coolant Inventory The transient analyzed was inadvertent starttp of the high pressure core spray pump (feedwater flow control *ailure to maximum demand was covered in category 1).
- 6. Decrease in Reactor Coolant Inventory Inadvertent opening of a safety / relief valve (covered in category 1).
The analysis of continuous control rod withdrawal during power cperation was made with the three-dimensional BWR Simulator code described in NEDO-20953, "Three-Dimensieral BWR Core Simulator," which has been approved by NRC. The analyses of the other abnormal tperational transients were performed using the REDY computer code described in, General Electric Topical Report NEDO-10802, ,
" Analytical Methods of Plant Transient Evaluations for the General Electric
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Boiling Water Reactor." 15-5 e amme . e enman een annue en a
The transient resulting in the highest system pressure was a generator load , rejection without turbine bypass which resulted in a peak system pressure of 1193 pounds - per square inch which is below the allevable maximum pressure of 1375 pounds per square inch gauge. Note, however, that a more severe transient is used for the Americ So:1ety of Mecnanical Engineers sverpressure protection report. This is an inadvertent main steam isolation valve closure with fai,1ure of the position switch scram also assumed (see Section 5.2.2).
- L For transients categorized under Decrease in Core CocJant Temperature, the most severe transient is the loss of feedwater heating event with manual flow .
control. The resultar,t minimum critical power ratio reached is 1.08 and the peak vessel pressere is 300 pounds per square inch below the ASME code limit. For this transient, the applicant assumed a 100 degree Fahrenheit drop in feedwater temperature. However, a drop of 150 cegrees Fahranheit had occurred , at a dcmestic boiling water reactor as the result of an electrical component f ail ure. In response to a staff 'request for justification of the temocrature drop assumed, the appiteant stated that the MCoR would remain essentially unchanged cue to the fixed scram setpoint even though the scram would occur earlier in time because of the faster power increase with a 150*F drop. The applicant has assumed operation of a thermal power nonitor circuitry to initiate scram for this event. We therefore require that'the time constant of the thermal power monitor be included in the plant technical specifications anc that it be tested periodically Inadvertent safety / relief valve opening causes a decrease in reactor coolant inver. tory and results in a cild depressurization event which has only a slight effect on fuel thermal margins. Changes in surface heat flux are calculated to be negligiole indicating an insignificant change in minimum critical power ratio. Thus, tne transient is found to be acceptable. The effect of inadvertent safety / relief valve opening on suppression pool temperature is treated in Section 6.2. The applicant was asked to justify that coeration with partial feedwater heating to extend the cycle bey.ond the normal end of cycle condition would not . result in a more limiting change in minimum critical power ratio than that 15-6
obtained using the assumption of normal feedwater heating. The applicant indicated that analyses will be provided prior to operation in this mode if a decision is made to operate in this mode. Until such analyses are provided we will condition the license from operation in this mode. For transients categorized under Decrease in Reactor Coolant System Flow Rate, the most severe transient is that resulting from simultaneous trip of both recirculation pump motors. As the pumpr. coast down, the core void fraction increases, causing level swell in the reactor vessel apd a decrease in neutron flux. Turbine trip occurs because of high water level. The minimum critical power ratio remains well above the safety limit and there is a small increase in reactor pressure. Increased recirculation flow because of flow control failure or startup of an idle recirculation pump would result in reactivity anomalies. These events are not limiting transients and neither primary pressure boundary nor fuel damage criteria are exceeded. The transient which could cause unplanned addition to coolant inventory is the inadvertent actuation of the high pressure core spray system. The high pressure core spray system actuation has a small
'effect, because its flow is small compared to the recirculation flow. The transient has little effect on fuel thermal margins and on reactor system pressure. We agree with the applicant's analysis. (The discussion of the rod withdrawal at power should be inserted here)*
Of the transients listed under Increase in Reactor Pressure, the limiting transients are turbine trip without bypass and generator load rejection without bypass which result in minimum critical power ratios of 1.07 and 1.05 respectively. The applicant has proposed however, that the generator load rejection without bypass and the turbine trip without bypass transients be classified as infrequent events. The reclassification of these events has been under review by the ctaff .and has not been approved. We require that these events be categorized as moderate frequency
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events and therefore the operating limit minimum critical power ratio must be adjusted so that the safety limit of 1.06 is not violated by these transients. We will report on this in a supplement to this report.
- LPM to add 15-7
In analyzing anticipated operational transients, the applicant has t.aken credit for plant operating equipment which is not normally reviewed by us because it is not considered essential for safety. We have discussed the application of this equipment generically with General Electric. Based on these discussions, it is our understanding that the most limiting transient, aside from generator trip without bypass, that takes credit for this equipment . is the excess feedwater event. Further, it is our understanding that the only plant operating equipment that plays a significant role in mitigating this event (excess feedwater) is the turbine bypass system and the level 3 high water level trip (closes turbine stop valves). In ordef to assure an accept-able level of performance for WNP-2 , our position is that this equipment be identified in the plant technical specifications with regard to availability, setpoints, and surveillance testing. The results of our review of the Technical ' Specifications with respect to the level 8 trip and turbine bypass system will , be addressed in Chapter 16.0. - As noted above, the operating minimum critical power ratio was also based in ' part on the REDY model described in NEDO-10S02. During the staff review of NE00-10802, three turbine trip tests were performed at Peach Bottom 2. As discussed in Section 4.4, these test's indicated that the results obtained with the REDY model are nonconservative for some events. We therefore require that the 00YN model, which has been' approved by the staff, be used to analyze the following transients. - ee e M t e
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(1) gr_Thermja jimit Enluation (A) Feecuater ccAtroller failure - maximum demand, (b) Ganerator load rejection witho~ut bypass, (c) Turbine trip without bypass, and - (d) Main steam isolation valve closures. (2) For American 5ccietv cf Mechanical Engineers Overcressure Protection (4) Main steam isolation v&lve closure with positfon switch scram failure (main stets isolatisn valve closure with flux scram). We will review the resulti cf these calculations to verify the acceptability of the operating minimum critical power ratio. 15.2 Accidents Tne applicant analyzed a pump tha?t seizurs and pump shaft break accident. Reacter scraa is suf ficient te praclude violiting the safety limit minimum critical power ratio (1,06) and, therefore. no fuel damage occurs. The reactor vessel pressuri is maintaIr.ed below the specified limit throughout the event. Hewr.er, the analyses inciaded the use of non-safety arade ecuioment. in partic; par,theturbictbypasss/3ter., and the high water level trip. The applicant hos f6erarued adgif tional inforratien in regard to the effect of these svstems on thermai margins. It was stated thai failure of the bypass and the high water level trip would result in an increase in the difference of critical power ratios of 0.08 which would still result in a minimum critical power ratio above the saf'ety , limit of 1.06. This satisfies the criteria of 10 CFR Part 100 and is therefore acceptable. l 15.2.1 Anticicated Transients Without_ Scram Background . Anticipated transients withcut scram (ATWS) are event in which the scram - system (reactor trip system) is postulated to fail to operate as required. This subject has been under ganacit review by the Commission staff for several years. 15-9
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In December 1978, Volume 3 NUREG-0460, " Anticipated Transient Withuut hram for Light Water Reactors" was issued describing the proposed type of pl' ant modifications we believe are necessary to reduce the, rhk from anticipated
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transients with failure to scram to an acceptable , level. . Wa issued requests
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for the industry to supply generic analyses to confirm'the anticipated transients without scram mitigation capability described in Volume 3 of NUREG-0450, and subsequently we presented our recommendations en plant modifi-cations to the Commission in September 1980. The Commission will determine the required modifications to resolve anticipated tran ient j without scram concerns as well as the required schedule for implementation of such modifica-tions. WNP- 2 is subject to the Commission's decision in this matter. It is our expectation that the necessary plant modifications will be implemented
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in one to four years following a Commission decision on anticipated transients i without scram. As a prudent course, in, order to further reduce the risk from anticipated transient without scram events during the interim period before completing the plant modifications determined by the Commission to be necessary., we require that the following steps be taken: - (1) An emergency operating procedure should be develo' ped for an anticipated' transient without scram event, including consideration of scram indicators, rod position indicators, average power range flux monitors, reactor vessel level and pressure indicators, relief valve and isolation valve indicators, and containment temperature,fpressure and radiation indicators. The emergency operating procedures should be sufficiently simple and unambiguous to permit prompt operator recognition of an anticipated transient without scram event. (2) The emergency operating procedures should describe actions to be taken in the event of an anticipated transient without scram including consideration of manually scramming the reactor by using the manual scram buttons, tripping the feeder breakers on the reactor protection system pcwer ~ distribution buses, scramming individual control rods from the back of the control room pan.el, tr_ipping breakers from plant auxiliary power - l source feeding the reactor protection system, and valving out and bleeding l l 15-10
off instrument air to scram solenoid valves. These actions must be taken immediately after detection of an ATWS event. Actions should also include prompt initiation of the residual heat removal system in the suppression pool cooling mode to reduce the severity of the containment conditions; and actuation of the standby liquid contfoi system if a scram cannot be made to occur. Early operator action as described above, in conjunction with the recirculation pump trip, would provide significant protection for some ATWS events, namely those which occur: (1) as a result of common made failure in the electrical portion of the scram system and some portions,of the drive system, and (3) at low power levels where .the existing standby liquid control system capability is sufficient to limit the pool temperature rise to an acceptable level. { l 15-11 l
II.B.1 Reactor Coolant System Vents - Requirements: See Item II.B.1 of NUREG-0737,-November 1980 Discussion and Conclusions - In Amendment. to the Final Safety Analysis Report tN6 applicant described the _ ventingprovisions at WNP-2. The primary venting capability is provided by the 18 power operated safety / relief valves. Each of the safety relief valves is seismically and Class 1E qualified and the air supply to the eight valves which comprise the automatic depressurization system is seismically qualified. These valves can be manually operated from the control room to vent the reactor coolant. system. Emergency procedures undertaken to assure core cooling under accident conditions will at the same time result in system venting and hence no specific venting procedures have been provided. Positive position indication for each valve will be provided in the control room by an acoustic monitoring system as discussed under item II.D.3 of this enapter. Additional venting capability is provided v'a a reactor vessel head ve'nt valve and through operation of the turbine driven reactor core isolation ccoling system. No additional accident analyses have been provided as a result of a break in any of these vent lines because a more bounding complete steam.line break is pa'rt of the plant's design basis. This is acceptable to us. i l
II .K.1 IE Bulletins on Measures to Miticate Small-Break LOCAs and less of Feedwater Accidents II.K.1.5 Assurance of Procer Encineered Safety Features Functionino" Requirement: See IE Bulletin No. 79-08,' April 1979' Discussion and Conclusion The applicant indicated that valve positioning directives and test and maintenance procedures associated with engineered safety features are currently in preparation and will be completed before fuel loading. WewillIonditiontheoperatinglicense e a for verification that these procedures are completed prior to fuel loading and that that all the requirements of this item have been met. Review and Modify ( As Recuired) Procedures for Removina Safetv-Related II .K.1.10 Systems From Service ( And Restorina to Service) To Assure Operability _ Status Is Known Recuirement: See IE Bulletin 79-08 April 1979 Discussion and Conclusions The applicant has indicated that the procedures for removing safety-related systems from service are ,in preparation and will be completed prior to fuel loading. We will condition the operating license for WNP-2 for. verification l that these procedures are complete before fuel loading and that all reautrements of this item have been met. O k l
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II.K.3.3 Report Safety and Relief Valve Failures Promotiv and Challences Annually Recuirement: See NUREG 0626 Item B.14 January 1980
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Discussion and Conclusion The applicant has committed to meeting the requirements of this item and to prepare the associated administrative procedure prior to fuel loading. The WNP-2 Technical Specifications will require that safety / relief failures be reported within 30 days. This is acceptable to us. II .K.3.13 Secaration of Hich Pressure Coolant Iniection and Reactor Core Isolation Coolina System Initiation Levels Recui rement: See NUREG 0737 Item II.K.3.13, November,1980 Discussion and Conclusions At WNP-2 the high pressure core spray (HPCS) and RCIC are both initiated at low-water level Level 2. , WNP-2 does not employ a high pressure coolant injection (HPCI) system. - As a gene' c item, the possible separation of initiation levels for RCIC and HPCS was studied by General Electric for the BWR Owners Group. The results of this study were forwarded to us by letter dated December 24,1980 from O. B. Waters (BWR Owners Group) to D. G. Eisenhut (NRC). The applicant' has endorsed the conclusi,ons, of that s,tudy and taken the position that the proposed separation of RCIC and HPCS initiation is unnecessary for safety - consi derations . The study concluded the following:1)for rapid level changes associated with accident scenarios and severe transients, HPCS and RCIC i
initiation wculd be essentially simultaneous in that possible separation j distances could not preclude HPCS cha11enges; 2) for slow level changes due to small leaks or slow transients, adequate tine exists for manual initiation of RCIC by the reactor operator prior to HPCS auto-initiation; and 3) no significant reductions in thermal cycles is achievable by separating the setpoints, nor is a reduction in cycles necessary.
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With regard to automatic restart of the RCIC system on low water level, the applicant has stated that this modification will be incorporated at WNP-2. We conclude that for WNP-2 the separation of HPCS and RCIC initiation levels is unnecessars, With regard to automatic RCIC restart on low water level, we require installation of this modification prior to fuel loading, or, if qualified equipment is not available before fuel loading, we require installation during the first outage of sufficient duration after qualified equipcent is available. II.K.3.15 Modify Break Detection Loaic to Prevent Sourious Isolation of Hich Pressure Coolant Injection and Reactor Core Isolation Coolina System Reouirement: See NUREG 0737 Item II.K.3.15 November 1980 Discussion and conclusion The applicant identified a circuit modification to assure _ that trips initiated by signals from pressure instruments used to sense flow ~ l in the RCIC system are actua11f base'd'en continuous high steam flow. The - modification is a time delay relay which is to be added to the logic of the RCIC systens . This conceptual design is acceptable. We require installation of the
4 rrodification before fuel loading if qualified equipment is available, or if qualified equipment is not available, we require instal.lation during the first outage of sufficient duration after qualified . equipment becones available.- II .K.3.16 Reduction of Challences and Failures of Relief Valves --Feasibility Study and System Modification Reouirement: See Item II.K.3.16 of NUREG 0737, Novembers1980 Discussion and Conclusions The applicant stated his participation in the BWR Owners Group Study of this item which was forwarded to us by letter dated fiarch 31, 1981, from D. B. Waters (BWR Owners Gmup) to D. G. Eisenhut (NRC). The applicant's-position based on the study is' that further modifications to the WNP-2
. would not sicnificantly reduce the frequency of safety relief valve events. Submittal of the applicant's position meets the current requirements for this item, but we note that WNP-2 is subject to the results of our generic evaluation of the Owners' Group report.
I t II .K.3.17 Report on Outace of Emeroency Core Coolina Systems Requirement: See Item II.K.3.17 of NUREG 0737, November 1980. Discussion and Conclusions
'The applicant has comitted to reporting a sunmary of emergency core cooling system outage data annually. _
This is acceptable to us. e S
s II.K.3.18 Modification to ADS Logic Requirements: See Item II.K.3.18 of NUREG-0737, November 1980 Discussion and Conclusions The applicant is a member of the BWR Owners' Group which has completed the required feasibility and risk assessment study. The results of this study have been transmitted in a letter from GE to NRC, D. B. Waters to D. G. Eisenhut, including retaining the current design, were considered. The applicant has taken the position, that the current ADS logic design, with implementation of the symptom-ariented emergency procedure guidelines (EPGs), is adequate. It is the staff's position that the applicant provide logic modification which eliminate the need for operator action to depressurize the vessel for the case of a stuck open~ safety relief valve or outside steamline break (with failure of HPCS). Alternatively, the applicant may provide analyses of containment heatup rates which demonstrate that a high drywell pressure signal will be present at a time .early enough to preclude exceeding the criteria of 10 CFR 50.46 for these events. We will discuss resolution of this matter in a supplement to this report. l l . l I i
, II.K.I.22 Describe Automatic and Manual Actions for Procer Functionino of Auxiliary Heat Removal Systems When Feadwater System is Not Operable
, REQUIREMENT: See Item 3 of Bulletin 79-08, NUREG-0560, May 1979 Discussion and Conclusions The applicant stated that if the main feedwater syste6 is not operable, a reactor scram will automatically be initiated when reactor water level falls to Level 3. The operator can remote manually initiate the reactor core isolation cooling (RCIC) system from the main control room or if the operator takes no action, reactor water level will continue to decrease from boil-off until the low-low level set-point, Level 2, is reached. At this point, the main steam line will be isolated automatically, and the high pressure core spray (HPCS) system and the RCIC system will be automatically initi'ated to supply makeup water to the reactor pressure vessel. These systems will continue automatic injection until the reactor water level reaches Level 8, at which time the HPCS and RCIC systems are tripped. The high pressure core spray system (HPCS) will restart automatically once the high-level trip signal clears and a low-low level (Level 2) signal is received. The RCIC design is being modified to provide for automatic restart following an L8 trip and a subsequent low-low level signal (see Item II.K.3.13 of this chapter). - If the vessel is isolated, reactor vessel pressure is regulated by automatic or remote manual operation of the main steam relief valves which blowdown to the suppression pool. In this case, the suppression pool cooling mode of the residual heat removal system is used to transfer heat to the ultimate heat sink. This requires remote manual alignment of the residual heat removal system valves and startup from the control room of the associated service water system. Reactor vessel heat removal may also be acomplished while the vessel is isolated by opera-tor action to align the residual heat removal system for the steam condensing ' mode of operation. This also involves remote valve alignments and startup of the residual heat removal service water system. ( For the accident situations with the reactor vessel at high pressure, the high - pressure core spray system is used to automatically provide the required makeup fl ow. No manual operations are required since the high pressure core spray system
, will cycle on and off automatically as water level reaches Level 2 and Level 8, respectively. If the high pressure core spray system fails under these cond'i-o tions, the operator can manually depressurize the reactor vessel using the auto-matic depressurization system to pennit the low pressure emergency core cooling systems to provide makeup coolant. Automatic depressurization will occur
- if all of the following signals are present: high drywell pressure, Level 3 water level permissive, Level I water level, pressure in at least one low pressure injection system and the runout of a 120 second timer which starts with the coin-cidence of the other four signals. <
Based on the description provided and the applicant's comitment to modify the RCIC system for automatic restart as noted above, we find the response to this item to be acceptable
*Also see item II.K.3.18 for possible modifications to the ADS logic.
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II.K.3.25 Loss of Power to Pump Seal Coolers REQUIREMENT: See Item II.K.3.25 of NUREG-0737, November 1980 Discussion and Conclusions In Amendment to the Final Safety Analysis Report the applicant stated that the BWR Owners Group supplemental input on this itewis applicable to WNP-2. The supplemental input was forwarded to us by letter dated September 21, 1981 from T. J. Dente (BWR Owners Group) to D. G. Eisenhut (NRC). Data from tests on a WNP-2 Bingham pump were provided. The test conditions were representative of boiling water reactor operating conditions. Observed leakage was less than 5 gallons per minute for more than 5 hours. This is acceptable to us. l a f me e e e b
. -- - -- ._ - - - , - ,, .,- - - - . - - -- -w-- ---.. - - . -
s II.K.3.30 Revised Small Break Loss of Coolant Accident Methods Requirement: See Item II.K.3.30 of NUREG 0737 November 1980 Discussion and Conclusions - The applicant stated that it is cognizant of the emergency core cooling system model changes being considered by the reactor s vendor General Electric. In conjunction with the applicant s response to
- Item II.K.3.31, this is acceptable to us. .
II.K.3.31 Plant Soecific Calculations to Show Comoliance with 10 CFR 50.46 Requirement: See Item II.K.3.31of NUREG 0737 November 1980 Discussion and Conclusions J ihe applicant committed to provide plant specific analyses for WNp-2 if any model changes are made in accordance with Item II.K.3.30 of this section. This commitment is acceptable to us. We note that NUREG 0737 requires these analyses by January 1,1983 or one year after staff approval of the model changes. II.K.3.44 Evaluation of Anticioated Transients with Single Failure to Verify No Fuel Failure Requirement: See Item II.K.3.44 of NUREG 0737 Noverber1980 Discussion and Conclusions The applicant endorsed as applicable to \NRP-2 the results of the Owners' Group ~ study 'in this area. The Owners Group report - was submitted to us by letter dated December 29, 1980 from D. B. Waters (Owners Group) to D. G. Eisenhut (NRC). The evaluation states that the worst case
6-transient-with-single-failure combination for BWR/5 plants is the loss of feed-s water event with failure of the high pressure. core spray system. A stuck open relief valve was also considered in addition to the high pressurt core spray failure. The results of these studies indicate that the core remains covered during the whole course of the transient either due to reactor core isolation cooling system operation, or automatic or manual depressurizatien
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permitting low pressure inventory makeup. The operator action assumed in the analysis is to manually depressurize the vesse.1 to permit low pressure injection. Based on the results of the Owners Group Study and their applicability to WNP-2 we find the applicants' response acceptable for this item. II.K.3.45 Evaluation of Deoressurization with Other Ther Automatic Deoressurization System Recui rerent: See Item II.K.3.45 of NUREG 0737 November 1980 Discussion and Conclusions The applicant has endorsed as applicable to WNP-2 the results of the BWR Owners Group study in this area. The analyses assumed failure of all hirh l pressure injection systems but operability of all low pressure systems. The time at l which the operator is assumed to actuate the automatic depressurization system varied. The effects!of depressurization over a 10-minute interval and a 20-minute interval were compared to full blowdown case which is completed in 3.3. minutes. The key parameter studied in regard to vessel integrity was vessel fatigue usage. The potential for a , reduction in fatigue usage as a result of a longer blowdown period was examined rela- ! tive to the impact on core cooling capability. It was concluded that: ,
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(1) Vessel integrity limits are not exceeded for full automatic , depressurization system blowdown. 1 I
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(2) For slower depressurization rates (longer than the approximate 3.3 minute interval associated with the normal depressurization rate), there is little impact on vessel fatigue usage relative to that usage assignable to the full automatic depressurization system blowdown. (3) Slower depressurization rates have an adverse impact on com cooling capability except when the cperator initiates blowdown very early in the
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accident. The results also indicated that some improvement in core cooling capability was passible using a 10-minute ' blowdown period if the~ operator actuated the automatic depressurization system within 1-6 minutes after the initiation of the accident. During this initial tine period however, it was considered i more prudent to attempt to activate the high pressure injection systems in an effort to avoid use of the automatic depressurization system. Based on the above discussion, we conclude that no changes to the current mode of depressurization are necessary for WNP-2 at this time. We note, however. - that the applicant is subject to the results of our generic review of this item. Response to Michelson's Concerns Requi rement: See Item II.K.3.46 of NUREG 0737 Discussion and Conclusions
'1he
, appi,icant has endorsed . the General Electric Conpany's generic response to the Michelson concerns as applicable to WNP-2. ;
The generic responses were forwarded in letters from R. Buchholz (GE) to , D. Ross (NRC) dated February 21, 1980 and from'D. Waters (GE) to D. Eisenhut (NRC) dated January 31, 1981. We reviewed the generic responses for
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. .' l their applicability to WNP-2 and found them acceptable.
This satisfies our requirement for this item and is acceptable for licensing of WNP-2. - t e i e-
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L bE . j . .- 5 fid','[t 9 NOV 3 0 881 - Docket No. 50-397 MEMORANDUM FOR: A. Schwencer, Chief Licensing Branch No. 2, DL FROM: R. Auluck, Project Mar.ager Licensing Branch No. 2, DL , 1
SUBJECT:
FORTHCOMING MEETING WITH WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS) to o DATE & TIME: Decemberfand M,1981 8:30 AM LOCATION: G. E. Conference Room Bethesda, Maryland PURPOSE: Discussion of open items in the Instruments and Controls Systems Area PARTICIPANTS: NRC J. Knight and R. Auluck l WPPSS R. Nelson and Support Staff Original signed by R. Auluck, Project Manager Licensing Branch No. 2 Division of Licensing cc: See next page
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NRC FORM 318 (10-80) NACM OHQ OFFICIAL RECORD COPY usom. an-mm
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l N . 3 . 9 ? Mr. R. L. Ferguson Managing Director
,Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 cc: Nicholas Reynolds, Esquire Debevoise & Liberman 1200 Seventeenth Street, N. W.
Washington, D. C. 20036 Richard Q. Quigley, Esquire j
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Washington Public Power Supply System P. O Box 968 Richland, Washington 993S2
. Nicholas Lewis, Chairman .
Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Wahington 98504 Roger Nelson, Licensing Manager Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352 , Mr. O. K. Earle, Project Licensing Supervisor Burns and Roe, Incorporated 601 Willians Boulevard Richland, Washington 99352 Mr. Albert D. Toth U.S.N.R.C. Resident Inspector WPPSS-2 NPS P.O. Box 69 Richland, Washington 99352 0 9 e __ __ _ -- n------- -- =
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CATED NOV 3 0 tent Occhet File (s) I&E Region I . h I&E Region II Local PDR I&E Region III Branch Reading File I&E Region IV
.51C I&E Region V TERA TIC NRC Particioants:
E. G. Case D. G. Eisenhut/R. Purple T. Novak S. Varga s - T. Ippolito R. A. Clark J. F. Stolz (ORB!4) R. Tedesco B. J. Youngblood A. Schwencer F. Miraglia E. Adensam (LB!?) J. R. Miller bec: Applicant G. Lainas Service List D. M. Crutchfield B. T. Russell Branch Licensing Branch fio. 2 J. Olshinski R. H. Vollmer Project Manger R. Auluck R. J. Mattson . - S. H. Hanauer Licensing Assistant . Service T. Marley. J. P. Knicht W. Johnston (AD/ Materials & Qualif. Engr} D. R. fiuller P. S. Check W. E. Kreger L. S. Rut:enstein F. Schroecer M. L. Ernst ~ ACRS (16) 0!&E(3) OSD(7) Attorney, OELD J. LeDoux, I&E V. Moore B. Grimes O s
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1GV 3 0 ;531 Docket No. 50-397 MEMORANDUM FOR: A. Schwencer, Chief Licensing Branch No. 2, DL FROM: R. Auluck, Project Manager Licensing Branch No. 2, DL
SUBJECT:
FORTHCOMING MEETING WITH WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS) lb + 17 OATE & TIME: Friday, December)(,1981 9:00 AM LOCATION: G. E. Conference Room Bethesda, Maryland PURPOSE: Discussion of open items in the Power System Area (Mechanical) PARTICIPANTS: NRC R. Giardina and R. Auluck WPPSS R. Nelson and Support Staff TGnke- _ R. Auluck, Project Manager Licensing Branch No. 2 Division of Licensing cc: See next page 811130 r'un mDOCK 05000397 N @A
iGV '3 0 1991 Mr. R. L. Ferguson Managing Of rector
- Washington Public Power Supply System P. O. Box 968 2000 George Washington Way Richl and , Nasn.; n.1 33352 cc: Nicholas Reynolds, Esquire Debevoise & Liberman 1200 Seventeenth Street, N. W.
Washington, D. C. 20036 Richard O. Quigley, Esquire /
~
Washington Public Power Supply System P. O Box 968 Richland, Washington 99352
. Nicholas Lewis, Chairman .
Energy Facility Site Evaluation Council 820 East Fifth Avenue 4 Olympia, Wahington 98504 Roger Nelson, Licensing Manager Washington Public Power Supply System P.O. Box 968 Richland, Washington 99352 , Mr. O. K. Earle, Project Licensing Supervisor Burns and Roe, Incorporated 601 Williams Boulevard Richland, Wasnington 99352 Mr. Albert D. Toth U.S.N.R.C. Resident Inspector WPPSS-2 NPS P.O. Box 69 Richland, Washington 99352 e
'-{' ' METins tioTicg nisTRiouTt0N - -
DATED NOV 3 01981 Dccket File (s) I&E Region I
@Local P,0R I&E Region II I&E Region III Branch Reading File I&E Region IV fiSIC I&E Regien V TERA TIC NRC Particicants:
E. G. Case D. G. Eisenhut/R. Purple T. Novak S. Varca - T. Ippolito R. A. Clark J. F. Stolz (ORBf 4) R. Tedesco B. J. Youngblood A. Schwencer F. Miraglia E. Adensam (LB!?) J. R. Miller bec: Applicant G. Lainas Service List D. M. Crutchfield B. T. Russell Branch Licensing Branch tio. 2 J. 01shinski R. H. Vollmer Project Manger R. Auluck R. J. Mattsen . - -
- 5. H. hanauer Licensing Assistant M. Service T. Murley J. P. Knight W. Johnsten (AD/ Materials & Qualif. Engr)
D. R. tiuller P. S. Check W. E. Kreger L. S. Rucenstein F. Schrceder M. L. Ernst ~~ ACRS (16) OISE (3) 050 (7) Attorney, OELD l J. LeDoux I&E V. tecore B. Grimes i
i "k E"ItMG UOTice DISTRiouTICM - - CATED NOV 3 01981 1 6f4RC PDR I&E Region I , I&E Region II Local PDR - I&E Region !!! Branch Reading File I&E Region IV f! SIC I&E Region V TERA TIC NRC Particicants: E. G. Case _. . s D. G. Eisenhut/R. Purple g, ! ,. [ j ,,. . .. ,
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0EC 1 1951 Docket !!o. 50-397 ME!10RA!iDU!1 FO?.: Villian J, 91rchs Executive Oirector for Operations FP.P.!': Harold R, Denton, Director / Office of t'uclear Reactor Peculation
$1 d CCT: N/.'IFOR0 SfiS*i!C ISSUCS - l'0VCr:DED. 6 TELEGRK' FP.0'1 P0"EP.T FER7150':
This nemerander- is in response to Tom Peh*'s request that we lav9ut tSe MtC side of the issue, nossible solutions nnr1 a proposeq list of attendaes to rectinos cron0Ied by t'r. Ferquson, The Hany'ina Director of 1:ashinnton Public pouar Supnly Systu:, the 1.*!!P-2 aDplicent.
':?.T Side of Iseue The issues raised by !?r. Ferouson deal tasically rith two issues t19t '11 the r.ensciences review schedule for iP:P-2 may ba in s?ricus jaomrdy and
?) the Surply Systeti has nct been fully responsive to '!PC ceccerr.s.
With reqerd to tha first issue, en May 4,1901, the 5t:ff rMuestad additienal inforriation on ',I"P-2
- site aeology ara se f s-olocy. T8t e (;uestions dealt uith difficult suMeet "atters such as capable fault *nn, near field carthquake suars and oaveley,ent of a site srec{fic snecteun. The Sunply System has incorrerated -any of these isseres in a seisq1c r*obability study. Ubile an exrosure analysis is useful. Annen.ity A to 10 CFR 100 requires that sr,.'c of these subject ratters be dealt deter *inistically o ersirically. ,
On Povenber 17 and la intensive discussions of tha ocor.cjences questior.s innk place involving the aanlicant end a lerne w5er of consi6tants ani cur ralevant stP'f and the U.S. Geolcojcal Survey. The staff hrs indicatad that these disctissions were productive and very pointed. It is obvious tFat svery ef fort was expended by the l'? P-2 staff to rs31'e tHs meeting i success and these effor'.s will contribute totard alleMing us to cmplete nur review. Consecuently, based ucon the forennin1, we are confident that the revieir schedule for 11' P-2 with reoard to reosciences s5nuld r.ct be in serious .iennardy provided the arritcant continues in its efferts to provide us with timaly aM adequate inforation. In order to further acrelerate car review, we Nye also taken some additional actions. Un have acquired the assistance of Dr. "Hid F1&ons, an expert j in earthquake fault evaluotien. He aleno uith Geosciences Pranc5 personnel j i ' su ma)
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l - W. 1. circk.s OI will participate in a field trip 5.hich is planned in early DeceTher for the
- nurpose of completing the fault capability evaluations. We have made the l 0. S. Geological Survey aware of the need for its co'pleted rtview on schedule
- so that we are able to reet our SER date.
I The second natter deals with the responsiveness of the Supply Syste.1 to our concerns. As stated above we had neetinns with the Supply Systen. Tre 1 discussions vore creductive and 5:e believe will help us in ccmpletin, cirr i review on a tinely basis. 'r'hile the discussions resulfed in tre nead to obtain edditional infor-'13 tion fro : UPPSS, we are confident that it can and uill be sunnlied based o'; the discussions at the aeeting. A draf t Of i':for-nation eneded was reduced to writinti and trans-itted to the Supply System on fiovember 20 Uith the resolution of these !!?.C staff cencerns, up expect that t5n ?!?C staff will he in a cosition to Co~'plete its UUP-2 reviou in the ce9 sciences area. We are therefore car.fident that the Su'.' ply f ysten has enu been rescon';ive to nrevidinn us with the necessary infon atien. Pronnscr' list nf MC / ttencees if a reetina is requested by ffr. forcuson, I would propese that tne folle.iine coople from t'RR attend this rectino. H. R. Denton D. G. Eisenheit P. H. Vo11ner . P. Auluck 4 0@el 6cied by H. L tuten ,, 81arold P.. Senton, .hirecter Office of Diclear Reictor f.eculatior. 0
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5, , .5-j NRC BHDA . I NM 01 RICHLA\D,VA NOV 6, 1981 -
. TVX T108240415NRC BHDA' '
. NUCLEAR REGULATORY COMMISSION . , . -
VASHINGTON, DC * *
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N.4 cy . ATTENTIONE MR. BILL DIRCHS, . EXECUTIVE DIRECTOR FOR OPERATIONS . '- i , NRC
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SUBJECT:
NRC REV IEV 0F HANFORD SEISMIC ISSUES 68 ' ' 1 j .,. DEAR BILLS , -
. t RECENTLY, A MEMBER: OF THE! NRR STAFF CONTACTED DON BOUCHEY OF I .
;,:. ,-),;lg,.
MY, STAFF, 'AND INDICATED THAT 'THE GEOSCf.ENCES REVIEV SCHEDULE FOR i '. VNP-2 MAY BE IN SER10U.S JEOPARDY AND THAT THE SUPPLY SYSTEM ' 1
,. j HAS NOT BEEN' FULLY RESP'ONSIVE TO'NRC CONCERNS RELATIVE TO THE -
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HANFORD , SITE SEISMIC ISSUES. THE NRR STAFF SUGGESTED AND 'VE\'.
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' VE AGREED rTO'*HAVE TECHNICAL AND MANAGEMENT MEETING S ON ' THIS ' I - .
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c'. TOPIC-IN M)D-NOV'MBER.- E . I VOULD LINE.70 MAHE TWO POINTS TO YOU 'REGARDING THIFeRSSUE. ,
, ' .I F!RST,.I VOULD DISPUTE STATEMENTS THAT THE SUPPLY' SYSTEM'HAS N'/, '
, BEEN "UNRESP0fJSIVE'.' ON HANFORD SEISMIC ISS'UES,! AND I ASSU E .
T # W'
'. YOU VE IN' EfJD TO CONTINUE A' RESPONSI'BLE POLICY IN : THIS HE ARD.
,( IN. FACT,.SINCE iHE VNP-2 PSAR, THE SUPPLY SYSTEM HAS'COND11CT ' !
-}@
,. , ED EXHAUSTIVF, , MULTI-MILLION-DOLLAR PROGRAMS DIRECTED AT' E' M' i *"
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SOLVING. NRC CONCERNS ABOUT HANFORD GEOLOGY 'AND SE5 SMI CITYk ' . .
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,. ,' THE NRd, STAFF HAS' BEEN PROVIDED ' DATA FROM THESE PROGRAMS JD '*
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,, . AND ,, HAS BEEN HEPT INFORMED OF THIS VORE THROUGH NUMEROUS' ' ' .Ib .' ['
f.EETIN,GS. .THOUGH MUCH NEW DAHB HAS BEEF 4 GENERSTED , NONE'0F .; !.. ' 7 I" THIS INFORMATION LEADS US OR OUR CONSULTANTS TO CONCLUDE , ' ' l
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... THAT CUR , CLPRENT SEISMI C DESIGN BASIS IS! INADEQUATE. ' .I BELIEVE -
- HAhlfwUNREsP0Ns!.VENEsS" TO NRC bTAFF , QUESTION 31STINN k,,g -0;;"'! W o9dk$.I351')M4.(( '
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/WONESTYEIFFEREP'Cc,., IN ,JUDGEMENTT DETWEEN cur, CONSULTANTS ANDU ' l/D!5 i*-
g THE NRC REV.IEVERS 'ON THE APPROPRI ATE METHODOLOGY FOR INTER-0 '
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PRETATION OF THE HNAFORD GEOLOGICAL AND SEISMIC DATA.
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d , SECONDLY,,I AM CONCERNEDABOUTCCUTINUINGTOPERPETUATESUPPLYj' I ' '. 1 ..,,, g-
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l . ;'[' SYSTEM -[TUNDED GEOLOGY RESEARCH PROGRAMS 'AND ."VMAT IF" SORT ,,
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OF QTIONSVVZ CONTRIBUTE TO IMPROVED SAFETY OF OUR
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. PLANTS.I IT APPEARS YO US'T.
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- . HAT # ART OF THE DIFFICULT - 5 3 s RAPIDLY tCHANGIN3 NRC STApF REVIEVERS, 'RESULTING EACH.' TIME IN .'
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, A NEV SERIES OF QUESTIONS, MEETINGS AND CONCERNS. I BELIEVE *
i 3 ., . THAT .tHE 'QUESTI ON OF THE APPROPRI ATE SEISMIC DESIGN BASIS FOR ! 4 i5 2 h ' Y . ., W ;,
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a THE HMAFORD SITE IS AN ISSUE VHOSE RESOLUTION' REQUIRES SOUND,'l P'
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k,, ., CONSIDERED 3 JUDGEMENTS AND I AM HO,PEFUL THAT THE NRR ,MANAGEME.NT ',' i ' ' , h
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h l , CAN DEVOh'E THE NECESSARY ATTENTION TO RESOLVE TH!S ISSUE IN '*- ' IN A TIMfLY 4 .- . . L i. IN A TIMELY FASHIONG I t eFUPPORT OF OUR VNP-2 OPERATING LICENSE j .;,.j , REVIEV.
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. ,1 DECAUSE THE SEISMIC ISSUE'AT HNAFORD IS OF SUCH BROAD SIGNIF- d n .*4 i ii ICANCE, INCLUDING THREE SUPPLY SYSTEM PLANTS, THE ENTIRE DOE M .l ;
- ' ', ; . }iANFORD COMPLEX (BOTH' CIVILIAN AND MILITARY PROGRAMS)'AND I 'v' i '
l? !' ' h THE NESCO SKAGIT PROJECT, AND BECAUSE VE VANT TO MAINTAIN I . b !
-[ , ;
- j. OUR UNP-S REVIEV SCHEDULE, I VOULD LIHE TO DISCUSS.THIS' MATTER
] }} VITH - -
YOU PERSONALLY:AS.500N AS POSSIBLE FOLLOVING THE TECHNI '
, . h '
CAL ANDj< MAN'A3EMENT MEETINGS BETVEEN OUR ' STAFFS, IN MID-NOVI?.EER. "". o . 8
.3- THANH YOU FOR YOUR CONSIDERA170N CF THIS Issue. I LOOC F' l [
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s VARD TO CHATTING. 4 ON THIS MATTER IN A FEV VEEKS. . i t . i .- MR. ROBERT FEP.GUSON .
'- . i* ' '
- l' 1 l' ' . MANAGING ; DIRECTOR d. - * *
.' VASHINGTON PUDLIC POVER SUPPLY SYSTEM. i i
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, **, N., , ', 3040 GEORGE VASHINGTON VAY.
1 1 J RICHLAND,VA 99352
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I i I . " 2!RC DHDA - ' j ,a.,.7 !LAST
... .,,. ,c s,0r.,MES$ fAGE.TO MR..D1LL.DIRCKSJ .
NRC SHdULD ,.".';h?*..rdj.;.'T;.%...... BE* . ..
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f5CJ HAROLD l,bE!JT0!i,. DIRECTOR,NRR ' ' VI CT OR ! STELL. O, DEPU37Y EXECUTIVE 3DI AREFFCTOR FOR - - OPERATIONS, NRC , l s . t . . m ,THANK YOU *j - '
.%,9. . ,.NR C. BHDA ,
l ffir0I RICHLANDcVA..* NOV 6, 1 9 1t I , t l 7108240415NRC BMDA . i
!,TVX NUCLEAR REGULATORY Cot:MI'SSICN * .
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VASHINGTON, DC - -
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,g, g ATTENTIONS. MR. BILL DIRCKS, EXECUTIVE DIRECTOR FOR OPERATIONSx.: .:.~ '
3.' URC , N-p"$.n$&:.Qr':.!?Ce.;. ::m. . u. n .., , . . . d . . .. e .- .
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f.71..".5. n 5DBJECTs.' .' NRC JREVIEV ? OF' HANFORD S. EISMIC ISSUES . . . . 3 { .
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DEAR. BILLt .
%KCCNTLY, A! MEMBER OF THE NRR STATF
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INTERIM REPLY } } [*)/Il
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{ DITCAS FINAL REPLY OF: s O CHAIRMAN' . FILE LOCATION O EXECUTIVE DIRECTOR { f.~
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OTHER: ^ **'*>- DESCRIPTION O LETTER OMEMO O aEpORT tJOTHER SPECI AL INSTRUCTIONS OR REMARKS Ike. fort Sdsric Isucs Prepre I rap ac c to OtrcLs layte . est !JO sfJe of .4 f: sue ani pssitie sol::tte.s as n11 at s tre esed list er attenho to n.ca',lais :rs;45:3 Ly Terysses, CLASSIFIED DATA T/Peh l OOCUMENT/COpV NO. j CLASSIFICATION l 4 NUUBER OF PAGES CATCGORY t OsTAL REGist Rv NO: O Ns O RO O rRO ASSIGNED TO: DATE INFORMATION ROUTING LEGAL REVIEW D FIN A L O COpv l'eiI O F. . h2? I I / I/ '. } PIPC($ , A55tGNED TO: DATE NO dGAL OBJECTIONS D* STIR NOTIF Y:
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D vEs O Nol NRC PC RfA C2 EXECUTtVE CIRECTOR FOR CFERATIONS DO /.'OT RE/JOVE THIS CO*', PRINCIPA!. CORRESPONDENCE CONTROL e _}}