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#REDIRECT [[W3P86-3379, Forwards Response to Request for Addl Info Re Tech Spec Change Requests to Implement Bypass for nonsafety-related High Speed Generator Level Trip & to Support Cycle 2 Operation]]
| number = ML20215L476
| issue date = 10/23/1986
| title = Forwards Response to Request for Addl Info Re Tech Spec Change Requests to Implement Bypass for nonsafety-related High Speed Generator Level Trip & to Support Cycle 2 Operation
| author name = Cook K
| author affiliation = LOUISIANA POWER & LIGHT CO.
| addressee name = Knighton G
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000382
| license number =
| contact person =
| document report number = W3P86-3379, NUDOCS 8610290059
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 11
}}
 
=Text=
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                                                              +  P. O. BOX 60340 l OUISI&AN POWER          A LIGHT  / 317NEW BARONNESTREET ORLEANS, LOUISIANA 70160  +  (504) 595-3100
        $?ONEsY$
October 23, 1986 W3P86-3379 A4.05 QA Mr. George W. Knighton, Director PWR Project Directorate No. '7 Division of PWR Licensing-B Office of Nuclear Reactor Regulation Washington, D.C. 20555
 
==SUBJECT:==
Waterford SES Unit 3 Docket No. 50-382 Technical Specification Change Requests NPF-38-23, 40, 42 Additional Information
 
==REFERENCES:==
(1) W3P86-2163 dated June 24, 1986 (2) W3P86-3323 dated September 25, 1986
 
==Dear Mr. Knighton:==
 
By the Reference (1) letter LP&L requested a Technical Specification change (NPF-38-23) to implement a bypass for the non-safety related high steam generator level trip. In Reference (2), changes to the shutdown margin (NPF-38-40) and special test exception (NPF-38-42) Technical Specifications were requested to support Cycle 2 operation.
In subsequent discussions, your staff requested additional information concerning these changes. Enclosed please find our response:
Attachment 1 - NPF-38-40                                          '
i                    Attachment 2 - NPF-38-42 Attachment 3 - NPI-38-23 I      Should you require any further information please contact Mike Meisner at                        j l
(504) 595-2832.
Yours very truly, 8610290059 861023                          jA PDR  ADOCK 05000382                    g'd L Q' /
P                  PDR K.W. Cook Nuclear Support & Licensing Manager l      Enclosures cc:  B.W. Churchill, W.M. Stevenson, R.D. Martin, J.H. Wilson, L. Kopp (NRC/NRR) ,
NRC Resident Inspector's Of fice (W3)                                                M "AN EQUAL OPPORTUNITY EMPLOYER"                                    l\g\
l
 
Attachment 1 to W3P86-3379 Page 1 of 3 Additional Information License Amendment Request NPF-38-40 QUESTION:
In reference to proposed change number 40 to Technical Specifications 3.1.1.1 and 3.1.1.2, Shutdown Margin:
(a) Which of the measured RCS cold leg temperatures is used for T-cold?
 
===Response===
The lowest of the eight (8) safety-related cold leg temperatures are generally used to determine the Shutdown Margin requirements. However, the equivalent boron concentrations calculated per Operations procedure OP-902-090, Shutdown Margin, have sufficient conservatism to account for normally observed differences in the RCS T-cold indications.        Thus, any safety-related RCS T-cold value may be used for the calculation of required Shutdown Margin.
(b) How does the Shutdown Margin when all full-length CEAs are fully inserted account for the highest worth CEA fully withdrawn (stuck-out)?
 
===Response===
By definition, the Shutdown Margin assumes that the CEA of highest worth is always fully withdrawn (stuck-out).      When the actual Shutdown Margin is calculated at Waterford-3, one CEA is assumed to be stuck-out to be con-sistent with the definition. However, in the safety analyses supporting this Technical Specification change, credit was taken for all CEAs being fully inserted. If, for example, the stuck CEA was worth 1%, then a Shut-down Margin of 1% would mean that the core is actually subcritical by 2%
(i.e., 1% Shutdown Margin plus the worth of the stuck-out CEA) since all CEAs have been verified to be fully inserted. The safety analyses that support this change (zero power steam line break, boron dilution, etc.)
were initiated therefore, with the core subcritical by 2% since all CEAs were assumed to be inserted.
(c) Why does the 1% Shutdown Margin in Technical Specification 3.1.1.2 (Figure 3.1-0) extend to 400 F rather than 200 F as in Tech Spec 3.1.1.1?
 
===Response===
Above approximately 200 F, the Shutdown Margin requirements depicted in Figure 3.1-0 are a direct result of the all rods in configuration and the analysis of the RCS cooldown and resulting reactivity transients associated with Steam Line Break accidents at different (initial) RCS temperatures. As the initial RCS temperature decreases, the potential RCS cooldown and associated reactivity transient are less severe. This less severe RCS cooldown and more favorable reactivity transient result in a 1% Shutdown l
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Attachmint 1 to W3P86-3379 Page 2 of 3 Additional Information License Amendment Request NPF-38-40 Margin requirement at 400 F. This explicit analysis as a function of initial RCS temperature was not done for Cycle 1; hence, the Shutdown Margin determined at hot zero power conditions was conservatively extended to 200 F.
(d) Justify the linear interpolation over 100 F between the 1% and the 4.15%
Shutdown Margin values in Figure 3.1-0.
 
===Response===
As stated in the response to question (c), above 200 F the Shutdown Margin requirements are a direct result of reanalyzing the RCS cooldown associated with the Steam Line Break accident at various (initial) RCS temperatures.
The linear interpolation shown in Figure 3.1-0 bounds the Shutdown Margin requirements for initial RCS temperatures between 400 F and 500 F. That is, specific RCS cooldowns were analyzed at intermediate temperatures (between 400 and 500 F) and were found to require less Shutdown Margin than that required by the linear interpolation shown in Figure 3.1-0.
(e) Discuss the results of an inadvertent boron dilution event initiated from an initial subcriticality of 1% allowed by Figure 3.1-0.
 
===Response===
As discussed in the response to question (b), when all full-length CEAs are inserted a Shutdown Margin of 1% (as shown in Figure 3.1-0) means the core is actually subcritical by over 2% (1% Shutdown Margin plus the worth of the most reactive CEA). The inadvertent boron dilution with a Shutdown Margin as allowed by Figure 3.1-0 is described in Section 7.4.4 of the Reload Analysis Report. Under these conditions there is at least 15 minutes from the time an alarm makes the operator aware that an unplanned boron dilution is in progress until a loss of Shutdown Margin occurs. This is consistent with Standard Review Plan Section 15.4.6 and shows that sufficient time exists for the operator to identify the event and take the required action to terminate it.
(f) Discuss the results of the CEA Withdrawal event from a subcritical or low power condition assuming the Shutdown Margin values given in Figure 3.1-0.
 
===Response===
The Shutdown Margin required to preclude (or mitigate) the consequences of a CEA withdrawal event from a low power or subcritical condition is bounded by the Steam Line Break (at high temperatures). At lower temperatures, protection from the CEA_githdrawl event is provided by automatic removal of the CPC bypass at 10 power. Additional discussion of the CEA with-drawal event from a subcritical or low power condition is provided in Section 7.4.1 of the Reload Analysis Report.
NS41208
 
Attachment 1 to
  .                                                                W3P86-3379 Page 3 of 3 Additional Information License Amendment Request NPF-38-40 (g) How does proposed Technical Specification 3.3.1.2 affect the definition of Shutdown Margin given in Section 1.0?
 
===Response===
There will be no change to the definition of Shutdown Margin given in Section 1.0 of the Technical Specifications. That is, the Shutdown Margin requirements described in proposed Tech Spec 3.1.1.2 do not include the worth of the most reactive CEA.
l  NS41208
 
Attachm:nt 2 to W3P86-3379 Page 1 of 2 Additional Information License Amendment Request NPF-38-42 QUESTION:
In relation to proposed change number 42 to Technical Specification 3/4.10.3, Special Test Exceptions:
(a) Why must CPC trips be bypassed during physics testing?
 
===Response===
In order to measure CEA worths during Cycle 2 physics testing, it is necessary to insert and withdraw CEAs outside of the normally prescribed sequence. If the CPCs were not bypassed, they would generate out of sequence per,alty factors that would result in a reactor trip.
The intent of this proposed Tech Spec change is not to request permission to bypass the CPCs during physics testing since this is already allowed by Technical Specification 3.3.1 (Table 3.3-1, Notes c and d); but rather to credit the log power trip for protection against power transients initiated at low power levels instead of relying on reduced reactor trip setpoints in the linear power channels (which is currently required by Special Test Exception 3.10.3).
In order to bypass all 4 CPC channels without generating a reactor trip, it is necessary to increase the CPC operating bypass permissive bistable toavalue_g%and10bove between  10                    the_  ower level of thermal      where physics testing is performed power).                                  (usually
                                                                                        , the CPCs would automatically come out of bypass at a thermal  If this were notof power    dong 10 % and potentially trip the reactor. Since this same bistable also serves as the threshold value for bypassing the log power trip (i.e., it is the minimum thermal power below which the log power trip can not be bypassed), setting it above the log power trip setpoint (0.257% power per Technical Specifi-cation 2.2.1, Table 2.2-1) removes the possibility of bypassing the log power trip. Thus, by bypassing all 4 CPCs, as described in proposed change number 42, a type of " electrical interlock" is created that precludes the log power trip from being bypassed. Therefore, the log power trip will provide the necessary protection for increasing power transients precluding the need to change the high linear power setpoint.
It should also be noted that increasing the CPC operating bypass permissive bistable setpoint above the log power trip setpoint does not preclude the CPCs from performing a protective function. That is, if the thermal power were to exceed the bistable setpoint (without causing a log power trip),
the CPCs would automatically come out of bypass and, if necessary, trip the reactor.
Finally, proposed change number 42 inadvertently omitted the word "either" following LCO 3.10.3a and the word "or" following LC0 3.10.3b. The corrected page is included at the end of this Attachment.
NS41210
 
.                                                                  Attachment 2 to W3P86-3379 Page 2 of 2 Additional Information License Amendment Request NPF-38-42 (b) Discuss any possible transients, such as CEA withdrawal events, which may arise due to the removal of the CEA withdrawal prohibit when bypassing the CPCs.
 
===Response===
The CEA withdrawal event, or any increasing power transient, would result in a reactor trip when the core thermal power reached the high log power trip setpoint (0.257%) or, when the core thermal power reached the value at which the CPCs would automatically come out of bypass and, if warranted, trip the reactor.
(c) What low power transients rely on the CPC trips?
 
===Response===
The transients that are initiated from a low initial core power level (e.g., the CEA withdrawal event) credit the CPC variable overpower trip for protection. Since the CPCs automatically come out of bypass, they will be available to provide protection against any transients that are initiated from low power.
(d) Does the high log power level trip require a faster rate of power increase to trip as compared to the high LPD trip? If so, what is the effect of the fact that it may not catch rapid power increases as quickly as the CPC trip?
 
===Response===
The high log power level trip does not require any rate of power increase to trip the reactor. The high log power level trip setpoint is set at an absolute value of 0.257% thermal power per Technical Specification 2.2.1, Table 2.2-1.
(e)                      operator to reset the CPC operating bypass permissive What  reminds setpoint  to 10 thg% after testing?
 
===Response===
Once the low power physics test program has been completed the CPC operatjngbypasspermissivesetpointisreturnedtoitsnominalvalue of 10 % power per station procedure NE-2-003.
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>    s
> . .Y      SPECIAL TEST EXCEPTIONS 3/4.10.3 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.3 The noted requirements of Tables 2.2-1 and 3.3-1 may be suspended during the performance of startup and PHYSICS TESTS, provided:
a.
The THERMAL POWER does not exceed SX of RATED THERMAL POWER, and Bill)ef I b.
The reactor trip setpoints of the OPERA 8LE power level channels are set at less than or equal to 20% of RATED THERMAL POWER, or                    l f APPLI        ILITY:    During startup and PHYSICS TESTS.
ACTION:
With trip the the THERMAL reactor. POWER greater than 5% of RATED THERMAL POWER, immediately SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%
j      of RATED THERMAL POWER at least once per hour during startup and PHYSICS TESTS.
4.10.3.2 Each wide range logarithmic and power level neutron flux monitoring channel shall be subjected to a CHAMEL FUNCTIONAL TEST within 12 hours prior to initiating startup and PHYSICS TESTS.
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Attachm:nt 3 to W3P86-3379 Page 1 of 4 Additional Information License Amendment Request NPF-38-23 QUESTION:
(a) Technical Specification proposed change number 23 requests providing the plant operators with the capability of bypassing the high steam generator level reactor trip. It appears that credit is indirectly taken for this          >
trip in events such as the increase in feedwater flow (FSAR Sec. 15.1.1.2) by using it to set the maximum water level used as an initial condition.
Please discuss. The statement is made that even should the main steam line piping be postulated to rupture due to the water loading, the resulting event is bounded by the main stea,n line break event analyzed in the FSAR.
Justify this by providing the results of a main steam line break analysis occurring at the highest possible steam generator water level and comparing these results (e.g., minimum DNBR, return to power, containment pressure and temperature) with the results of the FSAR steam line break analysis. Also, confirm that the resulting containment pressure and temperature is less than the LOCA containment pressure and temperature. Discuss the effect of possible higher temperatures than previously considered in the vicinity of the break on equipment qualification.
 
===Response===
Proposed Technical Specification Change Number 23 was requested to assist operators in reducing reactor trips during low power operation when steam generator level is being manually controlled.          Although not noted in the change request submittal, use of the high steam generator water level bypass will be administratively controlled such that the bypass cannot be enabled above 20% power. These controls will ensure that the high steam generator level trip is available to maintain steam generator inventory below the trip setpoint for events occurring above 20% power.
In this context, a distinction can be made between high and low power scenarios and the means available to limit steam generator water level.
At power levels greater than 20% the proposed Technical Specification change has no effect on plant operations -- i.e. the high water level trip remains.
available as in Cycle 1. For power levels below 20% the high level trip may be bypassed. However, for this case additional mechanisms exist to place an upper limit on water level.      First, the main feedwater isolation valves (MFIVs) receive a safety-related close signal (independent of the feedwater control system) when the steam generator level reaches 89.5%
narrow range (compared to the high steam generator level trip setpoint of 87.7% narrow range which may be in bypass).        Second, when the proposed bypass is in use the operator will have manual control of the feedwater system and be directed by procedure to maintain steam generator level at approximately 68% narrow range. The combination of these two factors ensures that steam generator level at low power levels will not increase to the point of filling the main steam piping.
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Attachm nt 3 to W3P86-3379 Page 2 of 4 Additional Information License Amendment Request NPF-38-23 Response (Cont'd):
The NRC's question consists of three elements:
: 1. The results of a postulated rupture of the main steam line due to water loading,
: 2. A discussion of the events which may indirectly credit a maximum steam generator water level, and
: 3. Confirmation that peak containment temperature and pressure due to Item 1 remain below that for LOCA.
These elements are covered in the following discussions.
: 1. Main Steam Rupture Due to Water Loading As discussed above, for situations when the high steam generator level trip may be in bypass (i.e..less than 20% power) sufficient automatic and manual controls exist to ensure that the main steam piping does not become water filled.
: 2. Indirect Credit For Maximum Steam Generator Water Level At power levels above 20% the high steam generator water level trip will not be bypassed and, therefore, will not affect FSAR events analyzed at high power levels. Low power events are discussed below.
Section 15.1.1.2 of the Waterford 3 FSAR states, in a discussion of the Increased Feedwater Flow Event, that protection against high steam generator water level is provided by a high steam generator level trip. Although at low power levels this trip could potentially be bypassed, allowing the steam generator to fill beyond the high steam generator level trip setpoint, the results of this event will not be as severe as the Increased Main Steam Flow which is discussed in FSAR Sections 15.1.1.3.
For Cycle 1, the Inadvertent Opening of a Steam Generator Atmospheric Dump Valve (IOSGADV) Event was initiated at zero power and an initial steam gen-erator water level just below the high steam generator water level trip set-i          point in order to maximize secondary side water inventory and the radiologi-i          cal consequences.      If the high level trip was bypassed, the potential exists for the initial steam generator water level to be as high as 89.5% (narrow range), at which point the Main Feedwater Isolation Valves (MFIVs) receive a close signal. LP&L has performed an analysis to determine the potential impact of the increased steam generator inventory on the radiological consequences of the IOSGADV.        The analysis conservatively assumed a 2%
process error on the closure signal to the MFIVs and took no credit for steam generator internals (e.g., steam separators) displacing some of the water. This resulted in a maximum increase in the steam generator inventory of approximately 21,000 lbm.
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Attachm:nt 3 to W3P86-3379 Page 3 of 4 Additional Information License Amendment Request NPF-38-23 Response (Cont'd):
Assuming the additional steam generator inventory was at the Technical Specification limit for iodine activity (0.1 uCuries/ gram) and all of the inventory is released directly to the environment during the first two hours of the event, the resulting increase in the thyroid dose at the site boundary is approximately 0.31 rem. This compares with the calculated site boundary doses shown in the FSAR of 5.5 rem (for the IOSGADV with no loss of off-site power) and 6.0 rem (for the IOSGADV with a concurrent loss of off-site power). This increase is within the calculative uncertainties of the analysis and the total dose is well within the guidelines established by 10CFR100.
The steam line break events presented in Section 7.1.5 of the Reload Anal-ysis Report, Steam System Piping Failures, are bounding for the following reasons:
The minimum DNBR during a steam line break accident occurs for full power initial conditions when a loss of off-site power is assumed to occur coin-
                                    ~
cident with the reactor trip.            For this case (described in section 7.1.5a of the Reload Analysis Report), the minimum DNBR occurs well before the affec-ted steam generator empties and, therefore, the initial steam generator water level has no impact on the results.
The minimum post-trip DNBR (and maximum post-trip power excursion) occurs assuming full power initial conditions and a loss of off-site power at the time of reactor trip (presented in Section 7.1.5b of the Reload Analysis Report).        At full power the high steam generator water level alarm and reactor trip is available to limit the inventory.          This full power case is more severe than a zero power case with a higher steam generator water level due to a greater number of delayed neutrons which contribute to the post-trip power increase.
The maximum radiological consequences of a steam line break occur during an outside containment break from full power initial conditions and a loss of off-site power at the time of the reactor trip (presented in Section 7.1.5a of the Reload Analysis Report). The full power case is again more limiting than the case with the maximum steam generator inventory (zero power) because it results in more postulated fuel failures.
: 3. Peak Containment Temperature and Pressures Operation with the high steam generator level trip in bypass may occur only below 20% power.
The peak containment temperature and pressure result from a 7.4765 ft 2 steam line break at 75% power as shown in Figures 6.2-7a and 6.2-7b of the FSAR. Therefore, a steam line break at 20% power or less will result in lower peak containment temperatures or pressures.
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Attachment 3 to a
W3P86-3379 Page 4 of 4 Additional Information License Amendment Request NPF-38-23 QUESTION:
(b) Where in the FSAR is there a discussion of the Steam Generator Overfill event? Specifically where does it evaluate filling of the Main Steam Lines up to the Main Steam Isolation Valves? If the evaluation is not in the FSAR what is the source document containing the evaluation?
 
===Response===
The FSAR does not address a steam generator overfill event.                                                        As discussed in response to the previous question, sufficient controls exist to ensure that the main steam piping does not become water filled.
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