ML20004D715
| ML20004D715 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/04/1981 |
| From: | Colbert W DETROIT EDISON CO. |
| To: | Kintner L Office of Nuclear Reactor Regulation |
| References | |
| EF2-53454, NUDOCS 8106090678 | |
| Download: ML20004D715 (18) | |
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June 4, 1981 EF2 - 53454 il @
Mr. L. L. Kintner U. S. Nuclear Regulatory Commission Division of Project Management Office of Nuclear Reactor Regulation 7920 Norfolk Avenue Bethesda, Maryland 20014
Dear Mr. Kintner:
Reference:
Enrico Fermi Atomic Power Plant, Unit 2 NRC Docket No. 50-341
Subject:
Responses to NRC Questions and Requirements Please find enclosed Detroit Edison's responses to several NRC requests. These items will be included in a forthcoming FSAR amendment.
Item I CPB Q241.5 Seismic and LOCA Loads in Fuel Detroit Edison's response to this question is enclosed as Attach-ment 1.
Item 2 RSB H.II.B.2.4,2 CS Definition Detroit Edison will voluntarily amend page H.II.B.2-3 of the FSAR to define the Core Spray System as a low-pressure system, as requested at the April 23, 1981 NRC/ Edison meeting in Bethesda.
Refer to Attachment 2.
Item 3 RSB FSAR5.4.6, 5.4.7 Valve Categorizatior Detroit Edison's response to this item is enclosed as Attachment 810 6 09 0 ( f g
1 1 Mr. L. L. Kintner June 4. 1981 EF2 - 53454 Item 4 RSB FSAR 6.3 HPCI Pump Reliability i
Detroit Edison's response to this item is enclosed as Attachment 4.
Item 5 PSB BTP-PSBl Degraded Grid Voltage Detroit Edison's response to this Branch Technical Position is enclosed as Attachment 5.
Item 6 i
RSB FSAR 5.2.2 S/R Valve Maintenance Detroit Edison's response to this item is enclosed as Attachment 6.
Item 7 1
RSB FSAR 15 Recirculation Pump Coastdown Detroit Edison's response to this item is enclosed as Attachment I
7.
1 Item 8 RSB FSAR 6.3 Assurance of Filled ECCS Lines I
Detroit Edison's response to this item is enclosed as Attachment 8.
Item 9 RSB Draft Position ECCS Pumps NPSH l
5/27/81 Detroit Edison's response to this draft RSB Position is enclosed l
as Attachment 9.
t Item 10 ICSB II.K.3.21 CS/LPCI Modifications t
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Mr. L. ' L. Kintner June 4, 1981 EF2 - 53454 Detroit Edison was requested to amend page H.II.K.3.21-1 to state that HPCI will automatically restart, following manual termination by the operator, should low water level again be reached. Attach-ment 10 includes this proposed change.
Sincerely, afL,0Ef William F. Colbert Technical Director Enrico Fermi 2 RMB/kw Attachment i
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Mr. L. L. Kintner June 4, 1981 EF2 - 54,454 Page 4 bec:
R. M. Berg F. E. Gregor J. W. Honkala E. Lusis L. E. Schuerman A. E. Wegele Document Control
ATTMC&W L Era - 53 Vsv RESPONSE TO QUESTION NO. 241.5 (CPB)
In response to' question 241.-5 (Core Parformance Branch), trans-mitted via NRC letter dated February 18,1981 (R. L. Tedesco to 4
W. H. Jens), req sesting documentation of the combined seismic and LOCA load analysis for the f uel assembly, the following is provided:
The fuel assembly was analyzed for the seismic and LOCA loads,
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including the Annulus Pressurization Loads and the results were documenteJ via FSAR Amendmant 29 in April, 1980. The results are presented in FSAR Table 3.9-40, showing a significant safety margin of a calculated maximum acceleration of 1.04 g versus the allowable acceleration of 3.1.? g.
The method of analysis and load combinations are discussed in FSAR Section 3.9.1.5.6.
dj e 5/29/81 r
wcws~r 9-EFo 2-FS AR yyg Section 6.4 of NUREG 75/087 (Reference 2) provided the guideline 5
for modeling the SGTS effluent plume.
p H.II.B.2.4.2 Radioactive Systems The systems assumed to contain radioactive liquids in ude the high-pressure injection system, the core spray system, the reactor core isolation coolant system, and the residual heat removal sys-tem, as well as portions of the control rod hydraulic system, sample lines, and all piping and equipment in communication with the primary coolant system out to the second isolation valve.
A design review has been performed to ensure that no systems other than those mentioned above would become contaminated with post-accident primary coolant.
In particular, design corrections have been made to ensure that the reactor building sumps (which could contain postaccident primary coolant) would not be pumped out of the reactor building.
The radwaste system, therefore, would not be contaminated by postaccident sources.
' Systems determined to contain postaccident primary containment atmosphere are the drywell, the torus free air volume, the hydro-gen recombiner system, all piping and equipment connected to the drywell, and torus free air volume out to the second isolation valve.
The reactor building atmouphere is assumed to be contami-nated as a result of primary containment leakage.
Steam lines are assumed to contain the core release fractions for airborne
'h sources outlined in Subsection H.II.B.2.4.1.
It is assumed that
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these sources are restricted to the vapor-containing areas of the primary coolant system.
A design review shows that the gaseous radwaste system is not exposed to postaccident source terms.
H.II.B.2.4.3 Radiation Environment The determination of the total radiation environment at any loca-tion includes the consideration of all of the many potentially con-tributing sources.
The sources considered include the following:
a.
Direct radiation shine from the airborne and liquid radiation sources in the drywell and torus b.
Direct radiation shine from essential safety feature (ESP) equipment and piping circulating postaccident contaminated liquids or gases in the reactor building (e. g., RHR, HPCI, RCIC, CSS, and hydrogen recombiners) c.
Immersion in and inhalation of the airborne sources within the reactor building HVAC boundary, resulting in ganaa whole-body doses, beta skin doses, and thy-l roid doses due to iodine inhalation.
d.
Direct radiation shine of the reector building and i
ref ueling floor atmospheres to
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l H.II.B.2-3 Amendment 33 - March 1981
W RSB Draft Qunction of 4/23/81 j pp9egW g CH -53vsy QUESTION:
- 8. Valve Categorization (5.4.6, 5.4.7)
- We require that motor operated valves which isolate the residual heat removal system from the reactor coolant system, or the RCIC system from the reactor coolant system be classified Category A in accordance with Section XI of the ASME code.
Check valves performing this function are to be classified A/C.
RESPONSE
Table 1 (attached) contains the list of valves performing an isolation function between high pressure and low press-ure portions of systems connected to the reactor coolant system.
These valves will be incorporated into the ASME Section XI Pump and Valve Testing Program and categorized as Type A or Type AC.
These valves shall not be routinely exercised every three months during plant operation as requir=d by IWV-3410 because:
- 1. Such tests would remove one of the two barriers protecting the low pressure portion of emergency core cooling systems.
- 2. The operators on testable check valves can-not overcome thu force on the valve with reactor pressure on one side.
RSB Draft Qusation (Continusd)
Ptga 2 The testing program for these valves will be:
EF-4 Exercise valve and verify valve position during refueling and after valve maintenance prior to return to service in accordance with IWV-3300 or INV-3522-(b)
EF-2 Exercise valve (full stroke) for operability during cold shutdown mode as time permits but not more frequently than once every three months.
ET Measure the full-stroke time in conformance with INV-3410.
(Not for check valves.)
SLT-4 Seat leak test the valve prior to reaching power operation following refueling and after valve maintenance prior to return to service.
M.
L.
Batch 6/21/81
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TABLE 1 List of Pressure Isolation Valves s
SYSTEM P&ID VALVE NUMBERS TYPE SIZE FUNCTION RHR
' 6M721-2083 Ell-FO15A, B Gate 24 Discharge to Recirc. System 6M721-2084 F050A, B Check 24 Discharge to Recirc. System F023 Globa 6
Discharge to Head Spray F022 Gate 6
Discharge to Head Spray F008 Gate 20 Suction from Recitc. System F009 Gate 20 Suction from Recirc. System F608 Gate 20 Suction from Recire. System CS 6M721-2034 E21-F005A, B Gate 12 Discharge to Core Spray Sparger
-FOO6A, B Check 12 Discharge to Core Spray Sparger HPCI 6M721-2035 E41-F006 Gate 14 Discharge to FW Line E41-F005 Check 14 Pump Discharge RCIC 6M721-2044 E51-(V8-2229)
Check 6
Pump Discharge E51-F013 Gate 6
Discharge +o FW Line i
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SAVANNAH RIVER QUESTivri NO. 1 (6.3)
The applicant must provide data to demonstrate HPCI pump reliability.
SPECIFIC CONCERN:
Basis for expected operating time of 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in Detroit Edison response to Q.212.67A.
RESPONEE:
The Figure of "500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />" for HPCI Pump was derived from the following:
Reference HPCI Pump Purchase Specification Data Sheet 21A9243AR, Rev. 4, Para. 4.3.1, "The unit will be tested once a month and may undergo several real starts during its 40-year lifetime." Surveil-lance tests are performed once a month, times 12 months, times 40 years equals 480 tests, plus a possible 20 real starts equals 500 operating hours.
dje 6/4/81
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j ff> - SM.SY EF2 FSAR BRANCH TECHNICAL POSITION PSBl - Adequacy of Station Electrical Dis-tribution System Voltage
RESPONSE
Fermi 2 has committed to install a second level of undervoltage relaying that addresses the concerns of the subject Branch Technical Position.
Specific features of the design are outlined below.
1.
The undervoltage relays are set in accordance with design calcula-tions to preclude damage to Class IE equipment. A time delay setting was chosen to avoid operation of the relay for motor starting conditions.
2.
Alarm relaying has been provided to alert the operatsrs that a / cod Vcitay c ' condition exists. The setpoint of the alart. relay is above that of the degraded grid trip setting. This wras done to give the operators advanced indication of system degradation.
j italsoeliminatesanypossibilitythatsetpointdriftwould permit the trip function to be actuated ahead of an alarm. It does not in any way affect the time delay of the degraded grid relaying.
3.
The time delay for actuation of the degraded grid undervoltage relay has been selected to be as short as possible without en-countering spurious trips due to motor statring.
E 4.
The degraded grid voltage protection system at Fermi 2 meets all applicable requirements of IEEE Standard 279-1971 " Criteria for Protection Systems for Nuclear Fower Generating Stations," as outlined in BTP PSBl.
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EF2 FSAR Page 2 5.
Upon loss of of fsite power, the emergency diesel generators start and, upon achieving synchronous speed, the automatic sequencer begins to sdd loads as required.
If a safety injection actuation c'enal is received, the sequencer will automatically shed all loads from the emergency diesel generators. The sequencer will then begin adding engineered safety feature equipment as needed to mitigate the consequences of the accident. The degraded grid relaying is not designed to operate during sequencer operation.
(Refer also to Question 222.33 of the Fermi 2 FSAR, Appendix E).
6.
The Class IE buses have been analyzed for all anticipated oper-ating situations. Refer to Chapter 8 of the Fermi 2 FSAR for a description of the Class IE Distribution System.
7.
Measurements will be made prior to full power operation to verify the Class IE buses analysis techniques employed.
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eTTACttrv1&vr-G IEra-ss ysy SAVANNAH RIVER QUESTION NO. 4 ADDITIONAL QUESTION: It was stated that up to 60 cycles are allowed between overhauls. Can GE verify this?
RESPONSE
The 60 cycle test described i;nder " Life Cycle Test" in response to Q.212.174 was to verify the ability of the design to meet its performance objectives. This test was not a life duration test to establish period between the overhauls-Response to 212.E 5, last sentence, commits to a 2 year maiatenance period. Based on operating tiperiaccejabout 45 SRV actuations per average year are predicted for the valve that actuates most frequently. The other SRV's would cycle much less frequently; less than 10 actuations per average year is predicted. These predicted actuations should be compared with the life cycling testing described in response to Q.212.174 when establishing the period between overhauls. The period between overhauls will also be a function of the problems encountered with the SRV's both during operation and during overhauls. A two year period between overhauls appears reasonable, initially. I dje 6/4/81
A77N6MM 7 M -SS45Y SAVANNAH RIVER QUESTION NO.13 A) In the applicant's proposed preoperational test of the recircu-lation system, the coastdown characteristics of the recirculation pumps apparently would be obtained by tripping the breakers of the drive motors for the motor generator sets. This would result in a slower coastdown which would not represent the ATWS recirculation pump trip in which circuit breakers open the genera-tor field. We require that the preoperational test of recirculation pump coastdown include a test involving the ATWS trip to verify that sufficiently conservative values of pump inertia were used in the analyses of the consequences and mitigation ef fects of this type of trip. The time delay for recirculation pump trip should also be verified during the test. B) What is the Level-2 trip for? (For NPSH protection only?)
RESPONSE
A) Detroit Edison Company will perform an RPT pump coastdown test to verify that suf ficiently conservative values of pump inertia were used in thr analyses where the trip occurs. The time delay for the trip will also be confirmed. B) The primary function of the low-water level (L2) trip of the recirculation pumps is to provide the NPSH protection for the recirculation pumps. l dje 6/4/81
~ .n gmcI Svy-f epa -nel Non-WI Coen Item 49 Assurance of Filled ECCS Lines (6.3) r 2e static head keep full system for the PCIC and HKI injecticn lines is 8 not sufficient. We require provisions such as those used to maintain the low pressure ECCS injection lines full to be rode for the HKI and FCIC discharge piping. In addition, the applicant must provide for incorporaticn into the technical specifications a schedule for periodic high point venting of all the ECCS injection lines and the PCIC injection line. Respor.g he PCIC and HPCI systems are normally lined up with the purp taking suction frce the condensate storage tank. All valves between the storage tank and ~ the first isolation valve en the p.mp discharge lines are open. This allcws ccrmu ication between the condensate storage tank and discharge line, through the BCIC and HPCI purps. De water level elevation in the condensate storage tank ranges frca 6/f[esTto a mini =.:m of about 594 feet. Elevations of the first isolatico valves in the PCIC and HPCI discharge lines are about 537 feet. No portion of the HKI and PCIC picp suction or discharge lines is higher in elevatico than the minimum water level in the condensate storage tank-here-fore, the elevation head of the condensate storage tank will ensure that the discharge line remains cc:pletely filled with water up to the isolation valves. The discharge lines ccnnect to the tottcm of the feedwater line. Therefore, the remainder of the di: charge lines will be maintained full by feedwater flow. Vent and drain connecticns are incorporated at high and 1cw points in tne PCIC and HPCI piping. Before initial start of the PCIC and HPCI system, the discharge lines are checked to make sure that they are filled with water by manually venting the high-Ecint vents to avoid trapped air pockets in these N' lines. retailed filling and venting procedures are included in the Cperating Pro- ~ cedure Manual and the precperational test peccedures and checks. Strict ad:nin-istrative procedures for.anually venting the discharge lines provide the main basis of assurance that these lines are full of water. P00ROREW1 ~
/ /4 / I Periodic high-point venting of all the KCS injection lines and ICIC injection line is already a part of the plant technical specification re-e quiremdnt. According to the plant technical specification, high-point ) venting cf the discharge lines is required once every 31 days to assure that the discharge lines on the ECCS and ICIC are full of water. l MKD:sm 5-18-81 i W y O 9 O e O e 9 e O t e 9 e* i s N t f J e w O g j I
ATTACW W f GF9-S]WG June 3, 1981 NRC QUESTION Calculations of NPSH available to ECCS pumps in BWRs are normally provided with reference to the pump suction. We are concerned that under certain post acci-dent conditions the potential may exist for damage to ECCS pumps from cavitation because of local flashing in the system suction lines. The potential can result for example from local elevation changes in the piping runs. Calculations of NPSH available at the pump suction may erroneously assume liquid continuity up to the point of pump suction. We require therefore that the applicants provide calculations demonstrating that all points in all safety related suction piping, the NPSH available is adequate to preclude local flashing under the worst postu-lated conditions.
RESPONSE
The pump suction piping isometrics for the HPCI, LPCI, and Core Spray systems were reviewed for possible local elevation changes. In all cases, the piping is routed so that the pipe elevation decreases from the suction source to the ECCS pumps, without any loops of increased elevation. S. Uema C. s' M' Ann -4
.___..vrm..~... N.WW /0 EF 2-FSAR H.II.K.3.21 Restart of Core Sdray and LPCI Systems on Low Level. O H.II.K.3.21.1 Statement of Concern Operator action could prevent the core spray and the low-preesure coolant injection (LPCI) systems from functioning when required, resulting in inadequate core cooling. H.II.K.3.21.2 NRC Position The core spray and LPCI systems may be stopped by the operator. These systems would not restart automatically on loss of water level if an initiation signal is still present. The core spray and LPCI system logic should be modified so that these systems will restart, if required, to ensure adequate core cooling. Be-cause this design modification affects several core-cooling modes under accident conditions, a preliminary design should be sub-mitted for NRC review and approval before making the actual modi-fication. The modification of system design should be made in accordance with those requirements set forth in Sections 4.12, 4.13, and 4.16 of IEEE 279-1971 with regard to protective func-tion bypasses and completion of protective action once initiated. Refer to NUREG-0660 and NUREG-0737 (References 1 and 2). H.II.K.3.21.3 Detroit Edison Position Detroit Edison endorses the BWR Owners' Group position that the p current BWR ECCS design, when coupled with rigorous and continu-Q ous operating staff training programs, represents the optimum approach to BWR safety. General Electric and the BWR Owners' Group have reviewed the modifications suggested in the NRC posi-tion, above. Their review concluded that the current emergency core cooling system (ECCS) design is adequate and that the pro-posed changes would have a negative impact on the overall safety of the plant. The negative impacts include a significant escala-tion of control system complexity, restricted operator flexibil-ity when dealing with anticipated events, and reduced system reliability. The conclusion that the current ECCS design is adequate is based on the comprehensive nature of BWR operator training, the emphasis placed in this training on reactor water level control, the Emergency Procedure Guidelines, the rela-and the ex-tively long time an operator has to correct errors, tent to which low reactor water level conditions are displayed and alarmed in the control room. As a result of this study by General Electric and the BWR Owners' Group, Detroit Edison has concluded that the above modifications suggested by the NRC should not be included in the Fermi 2 design. fcf/nN MtcbMWf kJvin n en, HACI bE bihr$' ) i utm6 ed cm w d b. 9 . h l)) $ M H.II.K.3.21-1 Amendment 33 - ch 1981 f
4 EF-2-FSAR H.II.K.3.21.4 References 1. U.S. Nuclear Regulatory Commission, NRC Action Plan Developed as a Result of the,TMI-2 Accident, NUREG-0660, May 1980; Revision 1, August 1980. 2. U.S. Nuclear Regulatory Commission, Clarification of TMI Action Plan Requirements, NUREG-0737, October 1980. / se t ~. h ~ H.II.K.3.21-2 Amendment 33 - March 1981
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