ML20004D543
| ML20004D543 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/04/1981 |
| From: | Colbert W DETROIT EDISON CO. |
| To: | Kintner L Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-2.D.3, TASK-2.K.3.16, TASK-2.K.3.22, TASK-TM EF2-53450, NUDOCS 8106090513 | |
| Download: ML20004D543 (26) | |
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Mr. L. L. Kintner U. S. Nuclear Regulatory Conunission Division of Project Management Of fice of Nuclear Reactor Regulation 7920 Norfolk Avenue Bethesda, Maryland 20014
Reference:
Enrico Fermi Atomic Power Plant, Unit 2 NRC Docket No. 50-341 Subj ect:
Responses to Miscellaneous NRC Ouestions and Requests for Information
Dear Mr. Kintner:
Please fir.d enclosed several Detroit Edison responses, as itemised below:
Item l_
ICSB II.K.3.22 Auto Switchover of RCIC
, provides a description of the proposed design changes to effect automatic switchover of RCIC per II.K.3.22.
This information will be included in a forthcoming FSAR amendment.
Item 2 MEB II.D.1 Safety / Relief Valve Tests Detroit Edison's response to the MEB position dated May 18, 1981 is contained in Attachment 2.
The FSAR will be amended as appro-priate.
Item 3 ASB QO20.26 Scram Discharge Volume Mods f h I i
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O Mr. L. L. Kintner June 4,1981 -
Page Two Detroit Edison's response to this question is enclosed in
- Attachment 3.. This information will be included in a forth-coming FSAR amendment.
Item 4 ASB QO20.28 CRD Roturn Flow-NUREG 0619 Detroit Edison's response to this question is enclosed in.
Please refer also to Item 5 below. This informa-tion will be included in a forthcoming FSAR amendment.
Item 5 ASB Q212.69A CRD Return Flow Line Detroit Edison's response to this question is enclosed in Attachment 5.
This information will be included in a forthcoming FSAR amendment.
Item 6 ICSB II.D.3 Valve Position Indication Detroit Edison has revised its Amendment 33 response to this item to address use of the S/R Valve tailpipe temperature monitoring recorder in conjunction with emergency operating procedures. Please refer to.
This information will be included in a forthcoming FSAR amendment.
Item 7 PSB Q222.31A, 32A Degraded Grid Voltage Detroit Edison's supplemental responses to these questions are enclosed in Attachment 7.
This information will be included in a forthcoming FSAR amendment.
Item 8 ICSB 7.7.1.3.3.1 ATWS "Monticello Fix"
Mr. L. L. Kintner June 4, 1981 Page Three
' Attachment 8 provides'a draft revision to FSAR page 7.7-23 to include-a statement relative to the "Monticello Fix" defined for ATWS mitigation. This information will be included in a forthcoming FSAR amendment.
Item 9 RSB II.K.3.16 Relief Valve Challenges' provides Detroit Edison's response to this NUREG-0737 itee. This information will be included in a forthcoming FSAR amendment.
Sincerely,
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L. E. Schuerman J. W. Honkala A. E. Wegele E. Lusis R. M. Berg F. E. Gregor Document Control
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I EH-S3 h~0 H.II.K.3.22.4 MODIFICATION A single failure proof automatic transfer of the RCIC suction lineup from the condensate storage tank to the suppression pool will be im, installed.
This change will utilize the existing redundant analog transmitter / trip
>:rl---4 M y unit tank level measurement which has been provided for the a
HPCI system.
A de-energize-to-operate relay logic derived from the level measurement channels will automatically open RCIC valves F029 and F031 (refer to FSAR Figure 7.4-1) on low condensate storage tank level.
The logic utilizes the full open position limit switches on F029 and F031 to initiate closure of F010 which is the valve in the pump suction connection to the suppression pool.
Panel status information is provided for the operator in the form of c
valve position indication and an alarm if the operator closes either suppression pool suction valve while the condensate storage tank level is low.
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Esa-534So De'roit Edison =
ENRICO FERMI UNIT 2 PROJECT ENGINEERING May 28, 1981 EF-2 #53414 To:
L.E.Schuerman Licensing Engineer 7.2 l W From:
D.F.Lehnert System Engineer, Sutj ect:
RESPONSE TO MEB POSITION ON NUREG - 0737, II.D.I.
Our response to the MEB position on NUREG - 0737, II.D.I. (attached) should be as follows:
"The preliminary safety / relief vahe (S/RV) operability test results demonstrating that the Fermi-2 S/RV-satisfies the acceptance criteria for operability is to be submitted to the NRC by July 1,1981. The final test report is to be submitted by October 1, 1981. The eval-uation of the associated discharge piping is to be submitted by January 1,1982. In addition, Edison will participate in develop-ment of any additional information applicable to the Fermi-2 docket with respect to $/RV operability and system functionability on a sch-edule consistent with that agreed to between BUR Qwners Group and the NRC staf f."
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5/86/61 i4Eb rusit. ion on II.D.1 In Appendix H to the FSAR, Amendment 33, ". arch 1981, Octroit Edisen responded In to the requirements of Action Plan Item 1I.0.1 as clarified by NUREG-0737.
this ree;:nse, the applicant comitted to participating in the BWR Chmers Group program for testing of safety and relief valves.
In addition, Detroit Edison is reviewing the CWR program description and scope to insure that it will be applicable to the Fermi-2 plant specific valves and piping.
The September 17, 1980 submittal from the BWR Owners Group referred to documentation, We are dates which are not completely consistent with those required by NUREG-0737.
continuing our review of both the September 17, 1980 submittal and the responses to NUREC-0737 received from BWR Applicants and Licensees. After we have completed our review of these submittals, we will arrive at a generic resolution regarding NUREG-0737 documentation submittal dates which will be applicable to all operating We require Detroit Edison to commit to orovide documentation in accordance reactors.
with this schedule for the Fermi-2 safety / relief valves and associated discharge piping.
b.Lekn h-In addition, on completion of the NRC staff review of the BWR Owners Group program, the staff may require additional information for pressures and temperatures 8
in excess of those provided in the low pressure test program described above,$
an a c/d i f-o b n a Detroit Edison is required to participate in development of 'me.information requested L
with respect to valve operability and system functionability for these temperatures and pressures on a schedule consistent with that agreed to between BWR Owners Group r and the NRC staff.
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CAP -SJ a/5'o RICUEST '!OR ADDITIOt&L INEORmTION ENRICO FETMI AICMIC POWER PIAVf, UNIT 2 AUXILIARY SYSTEMS BPANCH 020.26 Describe the means provided in the design of the scram discharge (4.6) system and verify that it meets the criteria enumerated in the Generic Safety Evaluation Report BWR Scram Discharge System, dated December 1,1980, and transmitted to you by NBC letter dated December 22, 1980.
0.20.26
Response
,The design of the fermi 2 scram discharge volume (SDV) will meet'the criteria stated in the Generic Safety Evaluation Report BWR Scram Discharge System, dated December 1,1980.
The plant specific items for Fermi 2 from this Generic SER t
are contained in this response.
The SER identified design deficiencies in the SDV are (a) inadequate hydraulic coupling between the SDV headers and D
the Instrument Volume (IV), (b) emplex vent and drain l
piping connections to the SDV, (c) failure mechanisms for the IV level instrumentation, and (d) failure of the control air system resulting in the potential inability to scram.
Items (a) and (d) are not applicable to Fermi 2 because adequate coupling is assured by the integral coupling of the twelve inch IV to the eight inch SDV header.
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'Ihe criteria of the BWR owners sub-group will be used in the design for items (b) and (c) listed in the previous paragraph.
The vent and drain valves will be made redundant and routed as
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dedicated lines. The instrumentation will be upgraded on each IV.
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' W.e instrumentation connections will be to the IV and not to the drain line. Each IV will have one (1) alarm, two (2) rod
- block and four (4) scram level instruments. The two (2) new scram level instruments will be diverse. These changes are being coordinated with General Electric.
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ef7TMCWr1W Y EfD -S29.To OUESTION 020.28 (4.6)
Describe the means provided in the control rod dri"9 system design to meet the criteria enumerated in NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking and verify that this design is in full compliance with this document.
Also a response to request 212.69A is necessary to complete our review.
RESPONSE
Question 212.69A requests our response to the letter l
s from D. Eisenhut, NRC, to R. Gridley, General Electric, dated January 28, 1980.
We believe the concerns expressed in this letter for the 251/BWR4 class of BWRs -
Fermi 2 is in this class - are the same as enumerated in NUREG-0619.
l The response to 212.69A shows Fermi. in in full com-pliance with all recommendations.
Please refer to l
the response to 212.69A for details.
M. L. Batch
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ATTJtCW&T" S EF.2 - S.;so Q212.69A Since the initial discovery of crackin3 in boiling water reacter (BWR)
(4.6) control rod drive return line (CRDRL) nozzles, General Electric (GE) has proposed a number of solutions to the problem. One solution GE has proposed is a system modification that involves. total removal of the CRDRL and cutting a;'d capping of the CRDRL nozzle. From your rerponse to Q212 69, Enrico Fermi-2 plans this modification. Address the applicable items and staff concerns specified in the letter from D. Eisenhut, NRC, to R. Gridley, GE, dated January 28,1980, on the subject of control rod drive return line (CRDRL) removal and capping CRDRL rozzles.
RESPONSE
Fermi _3 is.in conformance with all requirements in the referenced letter. Specifically:
Eqaalizing valves between the cooling water header a.
and the normal drive movement exhaust water header i.
have been incorporated, b.
Exhaust water headers have been changed to stain-less steel.
c.
Flow stabhizer loop is stainless steel and is routed directly to the cooling water header.
All modifications were constructed and inspected consistent with the applicable sections of the ASME B&PV code.
Detroit Edison agrees and commits to test for satisfactory i
system operation, return-flow capability equal to or in excess of the base-case requirement, and two pump operation.
The following reference is used for our base-case requirements:
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March 14, 1979 letter, G. G. Sherwood (General Electric) to V. Stello (NRC) and R. J. Mattson (NRC) re:
CRD Return Line Removal U
That letter specifies an injection rate for the Fermi 2 class of BWRs of 165 spm with one pump and two pump operation viability.
Detroit Edison also agrees and commits to establishing operating procedures for achieving this flow.
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,.4 6fd - 53.yS~o N.II.D.3 yalve Position Indi_<ar. ion H.II.D.3.1 Statement of Concern The operator needs a posi~tive indication of power-operated relief-valve and safety-valve positions to provide additional acnurance that the operator will correctly diagnose plant transients that potentially involve the opening of relief or safety vaives.
E.II.D.3.2 NRC Position Reactor coolant system relief and safety valves shall be provided
'with a positive indication in the control room derived from a re-liable valve-position-detection device or a reliable indication of flow in the discharge pipe, which meet the following requirements:
a.
The basic requirement is to provide the operator with unambiguous indication of valve position (open or
- closed) so that appropriate operator actions can be taken.
b.
The valve position should be indicated in the control room.
An alarm should be provided in conjunction with this indication.
c.
The valve-position indication may be safety grade.
If the position indication is not safety grade, 3 (f
reliable single-channel direct indication pcwe r e d from a vital instrument bus may be provided.f.tackup methods of determining valve position are available and are discussed in the emergency procedures as an aid to operator diagnosis of an action.
d.
The valve position indication should be seismically qualified consistent with the component or system
.I to which it is attached.
e.
The position indication should be qualified for its appropriate environment (any transient or accident that would cause the relief or safety valve to lif t)
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and in accordance with NRC Order, May 23, 1980 (CLI-20-81).
f.
It is important that the displays and controls added to the control room as a result of this requirement not increase the potential for operator error.
A human-f actors analysis should be per formed taking into consideration--
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The use of this information by an operator during' both normal and abnormal plant conditions A
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Integration into emergency procedures l 14y 1
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Other alarms during emergency and need for priori-tization of alarms H.II.D.3.3 Detroit Edison Position The Fermi 2 plant has a valve-position-indication system that uses one pressure switch on each of the 15 safety and relief valves.
The ersEEE=w tailpipe temperature-moni tor ing system its normal provides a backupb
~ monitors valve leakage as function.
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H.II.D.3.1 Modifications H.II.D.3.4.1 Desian Basis An in-containment, tailpipe-mounted, pressure-switch system provides status information via control room indicating lights of the relief valves and safety valves.
This system provides j
the following information to the plant operator during normal and abnormal operating conditions:
a.
Positive indication of valve position, including the
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stuck-open valve condition b.
Positive identification of the specific valve or valves that are open c.
Annunciation of the activation of the automatic depressurization system (ADS) in the control room By being provided with the immediate indication and annunciation of the valve opening and the identification of the valve, the plant operator can initiate recommended actions to control or rectify the situation.
The NRC has specifled in NUREG-0578 and NUREG-0737 (References 1 and 2) that components of the safety / relief valve (S RV) monitor l
system must be qualified for the appropriate environmental condi-tions to be experienced under normal and abnormal conditions of plant operation.
These environmental conditions include te=per-atore, pressure, and humidity, and also the seismic acceleration of the component or system to which the components of the SRV system are attached.
, The system instruments are quall'fied to IEEE 323-1974 and
.The power for this system comes from a reliable source that is not affected by the loss of offsite power,
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E7a. - s3 YSo t.: e E SUPPLEMENTAL ?.ESPOSSL TO lid' 222.3 t A Fermi 2 has co=nitted to install a secend level of undervoltage re-laying that addresses the above concerns. Specific features of the.
desi n are outlined below.
E 1.
The undervoltage relays are set in accordance with design cal-culations 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 operators that a low voltage condition exists.
The setpoint of the alarm relay is above that of the degraded grid trip setting. This was done to give-the operators advanced indication of system degradation.
It also _ eliminates any possiblity that setpoint drif t would 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 starting.
4.
The degraded grid voltage protection sys' tem at Fermi 2 meets all applicable requirements of IEEE Standard 279-1971 " Criteria for Protection Systems for Nuclear Power Generating Stations,"
as outlined in BTP.PSBl.
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5.
Upon loss of of fsite power, the emergency diesel g.:nerators star:
and, upon achieving synchronous speed, the automatic sequencer begins to add loads as required.
If a safety injection actuation signal 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 sequeneer 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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SUPPL.F.ME:;r.it hESPONSE This is to clarify the response to Ite: 2 2 2. L.i. Fermi.2 will per-for's testing and make actual cicasurements prior to full power operation as required. The measure =ents obtained from these preoperational tests will be use'd as data for the load flow program. New sets of voltages will then be calculated. The voltages obtained from the analytical techniques will'be compared to actual test results and other voltages calculated utilizing actual measurements as data.
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-;uT increase in recirculation fl w 6.P.;L -s'Ms o content of the acdcrator by increasingtemporarily reduces the void through the core.
the flow Jf coolant the reactivity of the core,The additional neutron moderation increases The increased steam generation rate increasescausing the reaut increase.
Oteam volume in the core with a consequent negative reactivity the effect, and a new steady-state power level is established.
When recirculation flow is reduced, in the reverse manner.
the power level is reduced with the turbine controls for automatic load following. Fi Each recirculation pump motor has its own motor-generator set for a power supply.
between the motor and generator of the motor-generator setA variab TO change the speed of the reactor recirculation pumpvar the changes the frequency and magnitude of the voltage supplied to the pump motor so.that I
the desired pump speed is attained.
The RFCS uses a demand signal from either the operator or the cnin plant d: mand signal is supplied to the master controller. turbine-genera The from the master controller adjusts the speed setting of the A signal cpced controller for each motor generator set converter.
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variable speed converter.The master controller signal adjusts each motor gen The caster controller signal is compared with the actual speed of the generator by the speed ccntroller.
the speed converter,The speed controller signal causes adjustment of resulting in a change of the generattr epned until the feedback from the generator equals the master enntroller signal.
,l Th3 reactor power change resulting from the change in reposition the turbine control valves.rceirculation flow causes the in cignal was a turbine load / speed error signal,If the original demand turbine control valves until the load / speed error signal isr the turbine r duced to zero.
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Etch of the two motor generator sets and its controls is id;ntical; therefore, only one description is given of the motor-generator set.
' curangement and rating of the motor-generator set. Figure 7.7-12 shows th g:narator set the motor-et any speed between approximately 19 percent and 96 percentca j
of. the drive motor speed.
The motor generator set is capable
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In order to mitigate the affects of an ATWS event, a provi-sion has been included in the Fermi 2 design to trip the re-circulation pump motor-generator field breakers using a s,Lecific logic philosophy defined as the "Monticello Fix This logic M includes reactor pressure and level r initiating signals which wSEE interface with redundant breaker trip coils in each MG set.
Suitable redundant time delay relays will be provided to maximize the single failure W withstand capability of the design.
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- II.K.3.16 Reduction of Challenges to SRV's (TMI-Open Areas)
Need final position based on the Omer's Group Study.
Response
Based on our evaluation of the reconnended nodifications-by the BWR Owner's Group Study, Detroit D31 son's final position is provided in revised FSAR sub-sections of H.II.K.3.16 l
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H.II.K.3.16.3 Detroit Edison Position In response to htREG 0737, item II.K.3.16, General Electric, on behalf of the BWR Owner's Group, has performed _a study of the feasibility and contraindications of reducing challenges to the-safety relief valves by various methods. This study reviews potential methods of reducing the lifihood of stuck open relief valvo (SORV) events in BWR's and estimates the reduction in such events that can be achieved by inplementing these methods. The resd ts of this study have been provided to the NBC.
Although' the NUREG-0737 position deals primarily with reduction of challenges to S/RV's, its clear intent is to reduce the incidence of SORV events. Reducing challenges is only one of three approaches to reduce SORV events. The other two are reducing the causes of spurious blowdowns and reducing the probability of S/RVs to stick open when challenged. All three of these approaches present feasible and effective opportunities for reducing the incidence of uncontrolled blowdowns via S/RV's.
Considering the proposed modifications by the Owner's Group, the goal of Detroit Edison is to identify feasible nodifications to the Fermi-2 design and operations which reduce the frequency of un-controlled S/RV blowdown to a fac rr of ten below the frequency experienced in BWR/4 bench mark plant (Bhw, W/3 stage Target Bock Valve).
Detroit Edison's evaluation of the proposed nodifications (Table 5.1 of Reference 3) indicates that the following nodifications already exist:
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Page 2
- 1. Two-Stage Target Rock Valves:
Use of two-stage target Bock Valves at Fermi-2 as ccxnpared to the plants with 3-stage Target Rock Valves, reduces the spurious blowdown events by 40 - 60% (Table 5.3, Reference 3).
- 2. Iow-Iow Set Relief or Equivalent Manual Actions:
Some BWR plants are equipped with a "Iow-Iow Set" design feature which changes t.he set points of selected SRVs following the initial opening of a number of S/RVs. This assures that following the initial pressurization the pressure will be relieved by the "Iow-Iow Set" valve alone, and the remaining S/RVs will not experience any subsequent actuations. Fermi-2 plant is not equipped with this design feature. However, the BWR Emergency Procedure Guidelines call for the equivalent manual action, wnich is part of Fermi-2 Emergency Procedure Guideline and it is estimated that a 23 - 62%
reduction in S/RV challenges can be achieved by implementing the equivalent manual action (low-low set relief).
According to the BWR Owner's Group evaluation these modifications (which already exist on Fermi-2) are equivalent to a reduction in S/RV challenges by a factor of aLmst ten (Table 5.1, Reference 3).
In addition to these existing nodifications, Detroit Edison will further reduce the SORV frequency at Fermi-2 by implementing a lowering of the RPV water level isolation set point for MSIV closure from Iavel 2 to Level 1.
This results in reduced S/RV challenges by eliminating isolation cycling of the S/RVs resulting from transients such as feedwater controller failure, trip of both recirculation pumps, and loss of feedwatc" flow.
H.II.K.3.1%3 - Detroit Edison Position' (Continued)
Page 3 Thus the two-stage Target Bock Valves and I.ow-Iow Set Relief equivalent manual action are two features which already exist on Fermi-2 design and contribute to the low expected SORV frequency.
In addition, Detroit Edison will lower the RPV water level isolation
. set-point for MSIV closure from Ievel 2 to Level 1, which will further reduce the number of SORV frequency. As indicated in Table 5.1 of Reference 3, these design features reduce the SORV frequency by.a factor of more than ten and hence the requirement of NUREG 0737 item II.K.3.16 is met for Fermi-2.
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- This modification will be coupleted before start-up followirg the first refueling'after NRC approval of the proposed change.
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BWR Owner's Group Evaluation of NUREG-0737, Item II.K.3.16,
" Reduction of Challenges and Failures of Relief Valves."
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