ML20136D575
| ML20136D575 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 07/06/1985 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML20136D511 | List: |
| References | |
| EOP-3.1-44, NUDOCS 8601060211 | |
| Download: ML20136D575 (22) | |
Text
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userr-t as.v. s-s' E0P 3.1-44
, Rsv. 12 G P1) 6 1985 JUL 3
OPER ATIONS REVIEW COMMITTEE APPROVAL Connecticut Yankee Emergency Operating Procedure N . Y' pera os ,
b/b ~
ABNORMAL FEED FLOW APPRoVEDB ' ATioy SU P E RINTEND ENT I
EFFECTIVE DATE
-AAW -
1.0 DISCUSSION Mv[' i 1.1 An abnormal feed flow condition will cause an automatic reactor and Turbine Trip as a result of:
1.1.1 Steam /Feedwater flow mismatch to Level on one steam generator.
1.2 Three examples of abnormal feed flow are considered in this procedure.
1.2.1 Feed line failure inside containment.
1.2.2 Feed line failure between containment anc feed station.
1.2.3 Feed line failure between feed station and feed pumps.
1.3 Do not override automatic actions of engineered safety features unless continued operations of engineered safety features will result in unsafe plant conditions.
1.4 Before opening any valves that have been closed by HCP initiation, verify by available instrumentation that the lines have integrity or is able to perform its intended function or not create a release path.
1.5 During a high energy pipe break in the contaument, close or check closed PORV isolation valves.
1.6 A feed line rupture inside the turbine building could effect both the main feed water control system and the aux. feed system. Except for the portion of the aux. feed water system that is in the terry turbine building in which case the feed to all of the steam generators could be started by opening MOV 35 from the main control board.
1.7 Feed line rupture near the atmospheric steam du=p valve could result in its failure in the open position adding to the consequences of the ,
accident. This will make it necessary to manually isolate the atmospheric steam dump valve locally.
8601060211 DR 851213 -
) ADOCK 05000y33 PDR Page 1 of 10
E0P 3.1-44 R;v. 12 (EOPI) Jg{ g jo85 ms
d 1.8 Reactor Coolant System VOIDING is possible during the course of this procedure.
A. VOIDING can be recognized by:
- 1. An unexplained increase in pressurizer level while trying to depressurize the RCS by pressurizer spray, power operated relief valve operation or other means.
- 2. An unexplained difficulty in reducing system pressure while operating with " solid" conditions in the RCS.
- 3. Indication of core thermocouple, reactor head or loop temperature readings equal to or greater than saturation temperature for a given RCS pressure.
- 4. Near zero reading of the subcooled margin monitor.
B. VOIDING can be overcome by:
- 1. Increasing RCS pressure by means of pressurizer heaters or charging rate to a pressure greater than saturation for a given RCS temperature.
- 2. Decreasing RCS temperature by feeding and steaming 3 steam generators, operation of core cooling equipment
_) or adjusting RHR cooling flow (if RHR is in service) to a temperature below saturation for a given RCS pressure.
- 3. Increasing RCS.flowrate which decreases the probability of hot spots occurring in stagnant or low flow areas of the RCS.
NOTE: During cooldown by natural circulation, care must be taken to cool the RCS slowly and uniformly so system hot spots are not created.
Voiding can occur in these hot spots when pressure is reduced.
1.9 On loss of feedvater or loss of heat sink, operators should consider using electric auxiliary feedpump to mitigate the consequences of an accident even though it is not Category I equipment.
2.0 SYMPTOMS 2.1 Decreasing steam generator level and pressure.
2.2 Reduction in steam generator feed pump discharge pressure.
2.3 Increase in containment pressure, humidity and sump level. (If break is inside containment.)
2.4 Loud audible noise and/or visible volume of released feed water and steam. (If break is outside containment.)
Page 2 of 10
,, E0P 3.1-44
, R v.12 (EOPI)
JUL 3.0 AUTOMATIC ACTION 61995 3.1 Reactor trip when stm. Gen. Lo level and steam flow feed flow mismatch occurs, or Hi sta. flow or L.P.S.C. actuates.
3.2 Turbine trip when reactor trips. >
3.3 The generator will separate from the grid 52 seconds after the turbine trips and 3091A and 3091B ACB's will open to separate 4160 V bus sections 1-1A and 1-1B from the main generator. 3TIA and 3TIB ACB's
'will close one second later to energize the bus sections from outside sources.
l 3.3.1 Two Reactor Coolant Pumps (1 and 3) are tripped on the bus !
transfer.
3.4 The feedwater control valves will travel to the full open position providing maximum feed flow to the steam generators.
t:
3.5 Core cooling will actuate on low reactor coolant pressure if pressurizer pressure decreases to 1700 PSIG.
3.6 Auto auxiliary feed will occur on 2 or 4 steam generator levels less than 45% on wide range or both main feed pumps off. j
.. 4.0 IMMEDIATE OPERATOR ACTION
-) '
i 4.1 Trip reactor. l I
4.2 Trip turbine.
4.3 Initiate core cooling, if pressurizer pressure <1700 psig.
4.4 Verify that all control rods are fully inserted by checking that all control rod bottom bistables are energized. Refer to Emergency Operating Procedure E0P 3.1-12, EMERGENCY B0 RATION, if more than one rod has failed to drop.
4.5 Evaluate main control board (HCB) indications to determine cause of abnoral feed flow. !
l Observe: Stm. Gen. levels and pressures [
Feed pump disch. pressures R.C.S. pressure and temperatures Pressurizer level Containment temperature, pressure and dew point Containment Sump Level 4.6 Verify auto auxiliary feed operation.
4.7 Verify non-effected steam generators are getting feed water flow. !
} I i
l Page 3 of 10 i
- ' E0P 3.1-44 R2v. 12 JUL 61085 4.8 Comumence emergency boration in accordance with Emergency Operating 7i Procedure E0P 3.1-12. EMERGENCY B0 RATION, to maintain adequate shutdown margin prior to reaching Tavg Temp. of 500*F.
4.9 If RCS pressure is less than 1700 psig, verify HPSI pump operation and stop All reactor coolant pumps (maintain seal injection flow).
5.0 SUBSEQUENT OPERATOR ACTION CAUTION: if the pressurizer power operated relief valves (PORVs) open at any time during this transient, verify that the valves reclose when RCS pressure drops below the PORV setpoint pressure. If either PORV does not reclose, attempt to isolate it by using the appropriate backup isolation HOV. If PORVs do not reclose and cannot be isolated go to E0P 3.1-4, I.oss of Coolant.
5.1 If core cooling is initiated verify containment isolation.
5.2 Stop any containment sump pumps that may be operating and place switches in trip pull out.
5.3 If indications are that the break is inside the containment e.g.,
containment pressure, temp. and des pt. increasing, increasing cont.
sump level, remove the affected reactor coolant loop from service.
] 5.3.1 Close the cold leg (Tc) loop stop valve and verify that the Reactor Coolant Pump trips automatically.
5.3.2 Stop feedwater addition to the affected steam generator by assuming manual control of auxiliary feed system and closing appropriate bypass valve.
5.3.3 Do not open HOV-35, but feed 3 uneffected steam generators via the by-pass valves.
NOTE: Do not add feedwater to a hot dry steam generator.
5.4 If the reactor coolant system T continues to decrease borate the systemtothecoldshutdownbor$nconcentration.
8 5.5 If indications are that the break is between the feed station and containment e.g.: Decreasing stm. gen. level and high indicated feed flow on the same stm. gen., reduced feed pump disch, pressure, and loud audible noise.
5.5.1 Close feed reg. valve on affected loop.
5.5.2 Close feedwater MOV isolation on affected loop.
)
Page 4 of 10
- E0P 3.1-44 Rev. 12
(* "')
JUL 61985
, 5.5.3 open MOV 35 to feed sem. generators.
t 5.5.4 Use feed line B/P A0V's to help control sem. gen. levels on non-effected generators.
5.5.5 Make cont. entry and isol. non-affected stm. generators from Aux. feed line in cont.
5.6 If indications are that the break is between the feed pumps and feed station e.g.: all steam generator levels decreasing, reduced or no feed flow on all sem. generators reduced feed pump disch. pressure and loud audible noise.
5.6.1 Shutdown main feed pumps.
5.6.2 Close feed pump disch. valves.
5.6.3 Shutdown condensate pumps.
5.6.4 Open MOV-35.
5.6.5 Close feed line MOV isolations.
5.6.6 If system is intact use feed line A0V B/P to control sem. gen.
level.
] 5.6.7 Close MOV-35 if B/P A0V's are used to control stm. gen.
levels.
5.7 Anticipate a residual heat buildup if the usual heat sink, steam dumping to the main condenser, is cut off by closure of the steam generator trip valves; open the atmospheric dump valves to maintain temperature and pressure.
NOTE: The possibility of voiding in the Reactor Coolant System exists whenever pressure is equal to or less than saturation pressure.
NOTE: Secondary safety valve and primary solenoid-operated relief valve operation may occur.
NOTE: If all reactor coolant pumps are tripped, operators shall monitor the degree of subcooling in the core by comparing core outlet temperature with saturation temperature for pressurizer pressure. Use the saturation curve (Attachment A to this procedure) and maintain pressure in the reactor core equal to or greater than " Saturation curve + 50*F safety Band".
Establish a flow producing a core LT greater than 10*F by steam dump / atmospheric vent operation in conjunction with
?
Page 5 of 10
- . . _ - ~
E0P 3.1-44 R::v. I2 (EOPI)JUL 61985 auxiliary feedwater flow. Monitor the potential for voiding
) by verifying a stable or decreasing core oT of less than 50*F.
Other instrumentation which can be used to monitor core conditions both during natural or forced circulation are listed in Attachment "B".
NOTE: If inadequate core cooling exists, refer to Section 8.0 of E0P 3.1-4, Loss of Coolaut.
5.8 In the event of a complete loss of feedwater and an eventual loss of heat sink, feed and bleed of the primary system must be initiated to remove decay heat from the reactor core.
The most likely event that could cause this situation is a loss of offsite power followed by a failure of the auxiliary steam generator feedwater system. Plant conditions at the start of this event would be pressurizer pressure rising to 2285 psia, and steam generator pressure at or below relief valve setting of 1000 psia. Core exit thermocouples should be rising to 575'F with steam generator levels decreasing.
Feed and bleed must be started before core exit thermocouples reach 575'F and before PORVs begin to automatically open at 2285 psia. All attempts should be made to restore feedvater supply to the steam l generators before initiating feed and bleed. Once PORVs are open, do not inject cold water into a hot, dry steam generator. To commence
} feed and bleed, ensure that at least one charging pump is on or start at least one high pressure safety injection pump, then open one PORV and its block valve and maintain in the open position. Do not cycle the PORY unnecessarily, especially if annunciator of low PORV air pressure occurs. Close letdown isolation (LD-MOV-200) if open, and maintain charging flow if power is available.
Pressurizer pressure will decrease and pressurizer level vill increase to 100%.
Check proper initiation of high pressure safety injection system when safety injection actuation occurs. Restart charging pump (s), if stopped, by safety injection actuation.
After 100,000 gallons have been pumped from RWST, establish long term high head recirculation by lining up RHR suction from containment sump and discharge to charging pump suction. Remain in this configuration until RCS pressure and temperature are low enough to go on shutdown cooling.
NOTE: Do not open two PORVs unless sufficient core cooling car.not be achieved with only one PORV.
O Page 6 of 10
E0P 3.1-44
, R;v. 12 (EOPI) 5.9 Safety Injection can be terminated if:
JUL 61985
)
(a) The reactor coolant pressure is greater than 1800 PSIG and increasing and, (b) pressurizer level is greater than 50% of span and, (c) the reactor coolant subcooled margin monitor indicates at least 50*F subcooled.
(d) The water level in at least one inserrice steam generator is in the narrow range span, or in the vide range span at a level sufficient to assure that the tubes are covered.
5.9.1 Stop LPSI and HPSI pumps.
NOTE: On emergency power neither charging pump will be running. It will be necessary to start one after stopping the low pressure safety injection pump.
5.9.2 Stop standby charging pump.
NOTE: If pressurizer pressure increases to 2100 psig start metering pump to supply RCP seal water and stop certrifugal charging pumps.
) 5.9.3 Place the following switches in the close positions:
SS-TV-955 SS-TV-960 SS-TV-950 SS-TV-965 CC-FCV-611 CH-HOV-331 LD-A0V-202 LD-A0V-203 LD-A0V-204 DH-TV-1841 DH-A0V-554 WG-A0V-558 5.9.4 Place control switches for the CAR Fan dampers in the " TEST" position. ;
5.9.5 Place containment sump pumps in TPO.
5.9.6 Trip RMS Channel R-12 (Vapor Container Gas) pump off prior to resetting containment isolation.
Page 7 of 10
. E0P 3.1-44 R v. 12 (EOPI)
JUL 6 1985 5.9.7 Reset high containment pressure relays (HCP A & B).
5.9.8 Reset Safety Injection WL relays.
5.9.9 Change operating charging pump suction from RWST to VCT and re-establish normal letdown and charging flow as per NOP 2.6-2.
5.9.10 Re-establish normal makeup and check that pressurizer controls are normal.
5.9.11 Stop diesel generators after 20 minutes if not needed.
CAUTION: If RCS pressure drops below the setpoint for safety injection or pressurizer level drops below 10% of span or the RCS subcooling is less than 50*F, MANUALLY REINITIATE SAFETY INJECTION, if not Auto initiated. The operator must rediagnose plant conditions and proceed to the appropriate emergency procedure.
5.10 If af ter securing safety injection and transferring the plant to normal pressurizer pressure and level control, the reactor coolant pressure does not drop below 1700 psig AND the pressurizer water level remains.above 10% of span AND the RCS is subcooled greater than 50*F, then:
5.10.1 Open one of the 12 valves in Step 5.9.3 and push the CIAS/HCP reset button behind the control board.
5.10.2 Close the safety injection and core deluge valves.
5.10.3 Place LPSI and HPSI and charging pumps in AUTO.
5.10.4 Reset the solenoids for the following valves and verify that the valves open:
DH-TV-1844 RM-TV-1848 WD-TV-lE46 WG-TV-1845 DH-TF-1843 LM-TV-1812 DH-TV-1842A S G-TV-1312-1 LM-TV-1811B SG-TV-1312-2 CC-TV-1831 SG-TV-1312-3 DB-TV-1847 SG-TV-1312-4 DH-TV-1842B MS-TV-1212 LM-TV-1811A MS-TV-1213 5.10.5 Place the following MCB switches in the normal position:
SS-TV-955 SS-TV-965 LD-A0V-202 DH-TV-1841 SS-TV-960 CC-FCV-611 LD-A0V-203 DH-A0V-554 SS-TV-950 CH-MOV-331 (([ LD-A0V-204 WG-A0V-558
)
Page 8 of 10
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2100 g'i
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'd 1900 ; .
- 1800 .
a U 1700 , El 1600 ,
- 1500
.c Q 1400 1300
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1200 TO NSE Tills GRAPH FIND 1100 THE AVERAGE TEMPERATURE 1000 ON THE BOTTOM TifEN COME 900 UP TO THE SAFETY BAND 800 CURVE AND MAINTAIN RCS PRESSURE AB0VE THAT LIMIT.
[00 gg IE Bulletin No.79-06A 500 400 300 200 m
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, E0P 3.1-44
", R:v. 12 (EOPI)
^"'"""'*
JUL 61985 Instrumentation available to aid the operator in monitoring pressure / tempera-ture relationships are:
Pressure
- 1. Pressurizer pressure - 3 independent channels that read out on the Main Control Panel. .
- 2. Reactor Coolant Loop #4 - 2 independent channels that read out on the Main Control Panel.
- 3. Drain Header Pressure - I channel that reads out on the Main Control Panel.
Drain Header Motor operated shutoff valve must be in open position.
- 4. Heisse Gage - I channel that reads out in the Primary Auxiliary Building.
Temperature
- 1. Loop T cold - 4 channels that read out Main Control Panel.
- 2. Loot T ave - 4 channels that read out on Main Control Panel.
- 3. Loop Delta T - 4 channels that read out on Main Control Panel.
j 4. Raactor Vessel Head Thermocouple - I channel that reads out on Main Control Panel.
- 5. Pressurizer Surge Line Thermocouple - I channel that reads out on main Control Panel.
- 6. Pressurizer Liquid Space Thermocouple - I channel that reads out on Main Control Panel.
- 7. Pressurizer Steam Space Thermocouple - I channel that reads out on the Main Control Panel.
- 8. Reactor Core Outlet Thermocouple - 40 channels that read out through an IBM 1800 process computer. Range up to 750*F.
Additional indications of void formation available to the operator are reactor coolant pump vibration alarms and erratic or unstable ammeter readings.
)
Page 10 of 10
Docket No. 50-213 B11913 Attachment C Haddam Neck Plant Heat Removal Systems and Components Outside Containment December,1985
Heat Removal Systems and Components Outside Containment Affected by High Energy Line Breaks Identification of High Energy Lines The identification of high energy lines was addressed by CYAPCO on June 29,1973 in response to the AEC Staff letter of December 18, 1972 requesting that CYAPCO address the consequences of postulated pipe failures outside containment.Section IV of CYAPCO's June 29 response listed " Feeding of Cool Water Direct to the Reactor Coolant System (RCS) through the Emergency Core Cooling System (ECCS)" (feed-and-bleed) as one of the three methods used to provide safe shutdown in the event of a pipe break outside containment.(l) While in many postulated pipe breaks more than one of the three methods would be available, the feed-and-bleed method was the only one available to achieve safe shutdown if a steam line break or feedwater pipe break at the main steam non-return valve area rendered both steam driven auxiliary feed pumps inoperable.
Extensive NRC reviews of CYAPCO's submittals on the topic of high energy line breaks resulted in NRC Staff approvals by letters dated March 29, 1974, January 17,1975 and July 14, 1977. These approvals concurred with CYAPCO's identification of high energy lines. A complete discussion of this issue is included in Attachment D.
As part of the SEP reevaluation effort, pipe breaks outside containment were reviewed and the results of this review were submitted by letter dated October 2,1981. The reevaluation included the following:
- 1. A comparison of the criteria used in the previous Haddam Neck high energy pipe break evaluation with current criteria;
- 2. A discussion on the available plant shutdown methods; and
- 3. An evaluation of break points which differed from previous pipe break studies.
By letter dated May 10, 1982, the NRC Staff accepted this evaluation for the areas where the " feed-and-bleed" method of safe shutdown would be utilized.
(1) While the actual scenario that would occur if this method were needed was t
not discussed in detail, it is clear from the 1967 discussion, as with subsequent analyses including that contained in Attachment A (accompanying this submittal), that the power operated relief valves would lif t resulting in blowdown to the containment and subsequent make-up from sources which would undoubtedly include the containment sump through the charging pumps.
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-- - - ,,.--r,v---,------- -,. - - - - , - . - - ,n- - - - - . ,. .-- - ,
y In summary the specific high energy lines whose rupture would expose the heat removal systems and components outside containment to a harsh environment have been previously submitted by CYAPCO and reviewed by the NRC Staff. No additional information on specific high energy lines is available.
Heat Removal Systems and Components Outside Containment Heat removal systems affected by high energy line breaks outside containment were identified by CYAPCO in our February 5,1975, October 31, 1980 and October 2,1981 submittals. By letter dated September 3,1981, CYAPCO identified the systems and components outside containment affected by high energy line breaks (HELB). The systems and components were addressed by plant location to determine which safe shutdown methods were available in the event of a HELB. A review of this submittal shows that only three areas credited
" feed-and-bleed" as the safe shutdown method available in the event of a HELB.
These areas are:
1.- Terry Turbine Building,
- 2. Building o'ver the Terry Turbine Building, and
- 3. Turbine Building The safe shutdown systems affected by these three areas are:
1 Main Steam g.
- 3. Service Water System s.
The components of the three systems located in the areas listed above are described in our October 31,1980 and June 1,1981 submittals on environmental qualification. We are presently reviewing this list for changes and will submit a list of components in a later submittal.
'f f
t C-3
Docket No. 50-213 B11913 Attachment D Haddam Neck Plant Chronological Summary and List of Docketed Correspondence Discussing " Feed-and-Bleed" December,1985
Attachment D Summary of Correspondence Regarding Staff Review and Acceptance of
" Feed-and-Bleed" at the Haddam Neck Plant The use of " feed-and-bleed" was first discussed by Connecticut Yankee Atomic Power Company (CYAPCO) in response to a question by the AEC Staff (Question 3B at p. 22) requesting information on what other means of decay heat removal would be available if auxiliary feedwater was lost. In its January 11,1967 response at p. 22, CYAPCO discussed this scenario and observed that as primary system temperature and pressure would rise, the pressurizer solenoid relief and safety valves would lift eventually resulting in blowdown to containment. The response smied that make-up would be effected through the charging pumps. An orderly shutdown would be achieved through manual control of the pressurizer relief valves and charging flow. This response was analyzed and found acceptable by the Staff in its response of May 12, 1967. A simplified piping schematic of the Haddam Neck Plant is attached to assist the reader in this summary.
The use of " feed-and-bleed" was next addressed by CYAPCO on June 29,1973 in response to the AEC Staff letter of December 18, 1972 requesting that CYAPCO address the consequences of postulated pipe failures outside containment.
Section IV of CYAPCO's June 29 response listed " Feeding of Cool Water Direct to the Reactor Coolant System (RCS) Through the Emergency Core Cooling System (ECCS)"
safe shutdown in (feed-and-bleed) as break the event of a pipe one ofoutside the three methods uses!)to provide containment.ll While in many postulated pipe breaks more than one of the three methods would be available, the feed-and-bleed method was the only one available to achieve safe shutdown if a steam line break or feedwater pipe break at the main steam non-return valve area rendered both steam driven auxiliary feed pumps inoperable.
Af ter an extensive review of the CYAPCO submittal which included at least two written sets of questions (e.g., letters from the Staff of October 1,1973 and January 24,1974) and one meeting (January 31, 1973), by letter dated March 29, 1974 from D. J. Skovholt to D. C. Switzer (and the attached SER), the NRC Staf f concluded that "[s]hutdown heat removal capability would always be adequate to remove core decay heat" in the event of a postulated " rupture of any high energy level piping system outside the primary containment." (SER attached to the March 29,1974 letter at p. 6.) In reaching this conclusion the Staff clearly understood that in the event of certain postulated accidents "cooldown could (1) While the actual scenario that would occur if this method were needed was not discussed in detail, it is clear from the 1967 discussion, as with subsequent analyses including that contained in Attachment A (accompanying this submittal), that the power operated relief valves would lif t resulting in blowdown to the containment and subsequent make-up from sources which would undoubtedly include the containment sump through the charging pumps.
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only be accomplished by feeding coolant directly to the reactor coolant system through the emergency core cooling system [i.e., feed-and-bleed]." at p. 4. The Staff further noted, however, that CYAPCO had committed to install an electrically driven auxiliary feed water pump (not safety-related) to provide "two independen.' methods of safe shutdown /cooldown for every postulated high energy line break assuming the unavailability of offsite power." The Staff's conclusions were reiterated in the December,1974 Safety Evaluation Report acccmpanying the Full Term Operating License.
On December 18, 1974, CYAPCO expressed its intent to substitute other modifications in place of those accepted by the Staff in its March 29,1974 letter to resolve this issue. (In making this proposal, CYAPCO was following Staff guidance which was later issued in a March 18, 1975 Mechanical Engineering Branch Technical Position No. 4, titled " Pipe Break Considerations in Plants Holding an Operating License Issued Prior to December 1972 - Augmented Inservice Inspection and Secondary Protective Measures.") By letar of January 17,1975, the Staff stated that it had no reason to believe that the alternative modifications would not be acceptable. Subsequently, by letters of February 5, March 13, April 4, May I and June 20, 1975; and May 10 and November 9,1976, CYAPCO formally proposed and answered Staff questions regarding modifications to be made in place of those previously accepted by the Staff to resolve the pipe break outside containment issue.
On July 14, 1977, the Staff issued Amendment 16 to the Haddam Neck Facility license. In the accompanying SER, the Staff provided the following discussion:
The licensee had originally planned to install an independent emergency shutdown panel and a motor driven auxiliary steam generator feedwater pump as a part of the actions to be taken to control the effects of high energy pipe breaks outside containment. On March 29, 1974, we issued Change No. 26 to Provisional Operating License No. DPR-14, (Facility Operating License No. DPR-61 was issued December 27, 1974) which approved the hardware modifications proposed by the licensee to provide an alternate means of safe plant shutdown in the event of a HEPB. Subsequently, the concept of preventing pipe whip and jet inpingement through the use of barriers, restraints and encapsulation sleeves and the use of augmented inservice inspection to assure continued integrity of high energy piping systems outside containment was approved on Millstone Units 1 and 2 and other operating reactors. In December 1974, as a result of material availability problems, schedule delays and increased costs, the licensee proposed a system of barriers, restraints and encapsulation sleeves and augmented inservice inspection as a revised approach to HEPB protection. By letter dated January 17,1975 we informed the licensee that the system of barriers, restraints and encapsulation sleeves could be substituted for the previously approved modifications pending our review and acceptar.ce of the detailed design. This revised approach to HEPB protection was approved for operating ,
reactors holding an operating license prior to December 1972 by D-3
Mechanical Engineering Branch Technical Position No. 4 which was issued March 18, 1975.
In approving the new modifications, the Staff accepted without reliance on an additional electrically driven auxiliary feedwater pump CYAPCO's use of feed-and-bleed as the sole method of removing decay heat in the event of certain postulated pipe breaks outside containment. Later, CYAPCO did install a non-safety related electrically driven auxiliary feedwater pump, whi-h provides even more protection regarding this issue than the Staff had determined was necessary. The existence of this pump was recognized as an element of the resolution of TMI Action Plan Item II.E.1.1 and the supporting SERs dated November 20,1981 and November 9,1982.
In an October 21, 1980 letter to all Operating Pressurized Water Reactors the NRC recognized that alternate ways may be available for removing decay heat following anticipated transients or accidents. The use of power operated relief valves (PORVs) as a means to " feed-and-bleed" was specifically mentioned.
By letter dated October 31, 1980, CYAPCO responded to a draf t Franklin Research Center (FRC) Technical Evaluation Report (TER) dated September 9,1980 on environmental qualification and noted that two independent trains of PORVs were available as an alternate means of removing decay heat.
The NRC responded by letter dated May 29, 1981 which transmitted an SER for CYAPCO's review. The SER contained an evaluation by FRC which questioned the use of " feed-and-bleed" in lieu of qualifying safety-related equipment outside containment following a high energy line break (HELB) accident.
By letter dated September 3,1981, CYAPCO addressed the concerns of using PORVs identified in the FRC TER. CYAPCO provided specific responses to the alleged deficiencies and concluded that in addition to the PORVs and the safety valves identified in the TER, the low temperature over pressure protection system is also available should both the PORVs or the safety valves fail to provide sufficient " bleeding" capacity. Sufficient " feeding" capacity is available under all potential reactor coolant system pressures. CYAPCO's response noted that emergency operating procedures describing the actions necessary to affect this method of decay heat removal had been implemented at the Haddam Neck Plant.
The revised FRC TER dated June 7,1982 was transmitted to CYAPCO in June 1982. The revised TER acknowledged CYAPCO's response to the previous concerns raised by FRC for areas outside containment. FRC's previous concerns on required instrumentation and the identification of all possible HELBs o :tside containment were resolved, but FRC continued to question the use of " feed-and-bleed" to obviate the need for environmental qualification of safety-related equipment located outside containment. The TER requested the NRC to resolve this issue.
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By letter dated December 13,1982, the NRC Staff Safety Evaluation Report for environmental qualification of safety-related electrical equipment addressed the disagreement between FRC and CYAPCO on the use of " feed-and-bleed" without seeking any additional information. The SER concluded that the qualification methodology and determination of environmental parameters for areas outside containment were acceptable. It should be noted that by letter dated February 11,1985, CYAPCO stated that the environmental qualification of equipment certification provided in response to Generic Letter 84-24 for the Haddam Neck Plant was based, in part, upon CYAPCO's reliance on the " feed-and-bleed" method of decay heat removal for certain high energy line breaks.
The use of " feed-and-bleed" was also raised in the Systematic Evaluation Program. By letter dated February 12, 1982, CYAPCO provided a Safety Assessment on SEP Topic III-4.C, Internally Generated Missiles. The assessment proposed to use " feed-and-bleed" (charging pump to feed and PORVs to bleed) as a safe shutdown method with respect to internally generated missiles. By letter dated May 10, 1982, the NRC Staff recommended that the use of the " feed-and-bleed" method should be addressed during the SEP integrated assessment.
By letter dated May 10, 1982, the NRC Staff accepted the use of " feed-and-bleed" as an available safe shutdown method with respect to SEP Topic III-5.B High Energy Pipe Breaks Outside Containment. The acceptance was based on CYAPCO's October 2,1981 safety assessment and February 5,1975 letters.
CYAPCO's October 2,1981 letter discussed the viability of " feed-and-bleed" using PORVs and charging pumps.
By letter dated August 2,1982, the NRC Staff accepted the use of " feed-and-bleed"(charging pump to feed and PORV to bleed) as an alternate safe shutdown path with respect to SEP Topic Ill-4.A, Tornado Missiles. The " feed-and-bleed" method reviewed was based on CYAPCO's February 5,1975 letter.
By letter dated September 17, 1982, CYAPCO provided a safety assessment on SEP Topic III-5.A, Effects of Pipe Break on Structures, Systems and Components inside Containment. The assessment discussed the use of " feed-and-bleed" using the ECCS to feed and the PORVs to bleed. It was emphasized that this is the least desirable method to remove decay heat. By letter dated October 12,1982, the NRC Staff requested more information on the " feed-and-bleed" method employed but did not take issue with its use. This information was provided by CYAPCO letter of December 17,1984.
The background historical discussion above clearly shows that CYAPCO has consistently docketed, explained and defended its position that " feed-and-bleed" is a method of mitigating the consequences of design basis accidents. We have consistently stated that " feed-and-bleed" is the third or least desirable method of decay heat removal. We believe that a fair reading of the Staff's correspondence reflects an acceptance of CYAPCO's position.
D-5
Chronological List of References D. C. Switzer letter to P. A. Morris, dated January 11,1967.
P. A. Morris letters to S. R. Knapp, dated May 10,1967 and May 12,1967.
A. Giambusso letter to D. C. Switzer, dated December 18, 1972.
D. C. Switzer letter to A. Giambusso, dated June 29, 1973 which included a special report entitled the Effects of a High Energy Piping System Breaks Outside of Containment.
D. 3. Skovholt let ter to D. C. Switzer, dated March 29, 1974 and the accompanying Safety Evaluation Report for High Energy Line Breaks Outside Containment.
December 18, 1974, D. C. Switzer to K. R. Golfer, High Energy Pipe Break Outside Containment.
Section 5 of the supplement to the SER accompanying the FTOL, dated December 27, 1974.
January 17, 1975, K. R. Coller to D. C. 'witzer, High Energy Pipe Break Outside Containment.
D-6
D. C. Switzer letter to K. R. Goller, dated February 5,1975 and accompanying report on the Effects of High Energy Pipe Breaks Outside the Haddam Neck Containment.
March 13,1975, D. C. Switzer to R. A. Purple, High Energy Pipe Break outside Containment.
Mechanical Engineering Branch Technical Position No. 4, dated March 18,1975.
April 4,1975, D. C. Switzer to R. A. Purple, liigh Energy Pipe Break Outside Containment.
thay 1,1975, D. C. Switzer to R. A. Purple, High Energy Pipe Break Outside Containment.
June 20,1975, D. C. Switzer to R. A. Purple, High Energy Pipe Break Outside Containment.
A. Schwencer letter to D. C. Switzer, dated July 14, 1977 and accompanying Safety Evaluation Report.
Franklin Research Center Technical Evaluation Report, dated September 9,
_1980.
D-7
D. G. Eisenhut to All Pressurized Water Reactor Licensees, date1 October 21,1980,
Subject:
Seismic Qualification of Auxiliary Feedwater
- 5) stems (Generic Letter 81-14).'
W. G. Counsil letter to D. M. Crutchfield, dated October 2,1981.
Franklin Research Center Technical Evaluation Report, dated June 7,1982.
W. G. Counsil letter to D. G. Eisenhut, dated October 31,1980 and Section B.7 of the Attachment.
D. G. Eisenhut letter to W. G. Counsil, dated May 29, 1981. The section of interest is Appendix H of the Technical Evaluation Report which discusses " feed and bleed" in relation to equipment located in areas outside containment and required for plant shutdown.
W. G. Counsil letter to D. G. Eisenhut, dated September 3,1981,
Subject:
Environmental Qualification of Electrical Equipment.
W. G. Counsil to D. M. Crutchfield, dated October 2,1981,
Subject:
SEP Topic III-5.B, High Energy Pipe Break Outside Containment.
D. M. Crutchfield to W. G. Counsil, dated November 20,1981,
Subject:
Auxiliary Feedwater System Evaluation - NUREG-0737, item II.E.1.1.
D-8 m
W. G. Counsil letter to D. M. Crutchfield, dated February 12,1982,
Subject:
SEP Topic III-4.C, Internally Generated Missiles.
D. M. Crutchfield letter to W. G. Counsil, dated May 10, 1982,
Subject:
SEP Topic ill-4.C, Internally Generated Missiles.
~
D. M.' Crutchfield letter ;o W. G. Counsil, dated May 10, 1982,
Subject:
SEP Topic III-5.B, High Energy Pipe Break Outside Containment.
D. M. Crutchfield letter to W. G. Counsil, dated August 2,1982,
Subject:
SEP l Topic III-4.A, Tornado Missiles.
W. G. Counsil letter to D. M. Crutchfield, dated September 17, 1982,
Subject:
SEP Topic III-5.A.
D. M. Crutchfield letter to W. G. Counsil, dated October 12,1982.
D. M. Crutchfield letter to W. G. Counsil, dated November 9,1982,
Subject:
Auxiliary Feedwater Evaluation - NUREG-0737, Item II.E.1.1.
D. M. Crutchfield letter to W. G. Counsil, dated December 13,1982.
W. G. Counsil letter to 3. A. Zwolinski, dated December 17,1984,
Subject:
SEP Topic III-5.A.
D-9 s
s .
D. G. Eisenhut letter to All Licensees of Operating Reactors and Applicants for an Operating License dated December 27, 1984,
Subject:
Certification of Compliance to 10 CFR 50.49, Environmental Qualification of Electrical
- Equipment important to Safety for Nuclear Power Plants.
W. G. Counsil letter to H. L. Thompson, dated February 11, 1985,
Subject:
Response to Generic Letter 84-24.
D-10
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