ML20207C903
| ML20207C903 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 12/24/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| NLR-N86183, NUDOCS 8612300275 | |
| Download: ML20207C903 (29) | |
Text
F p.
m Pubhc Service Electric and Gas Company Ctrbin A. McNeill, Jr.
Public Service Electnc and Gas Company P.O. Box 236 Hancocks Bridge, N.108038 609 339-4800 Vice President -
Nuclear December 24, 1986 NLR-N86183 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention:
Ms. Elinor G.
Adensam, Director Project Directorate #3 Division of BWR Licensing
Dear Ms. Adensam:
REQUEST FOR ADDITIONAL INFORMATION SINGLE LOOP OPERATION TECHNICAL SPECIFICATION HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Public Service Electric and Gas Company (PSE&G) hereby responds to the Request for Additional Information (RAI) contained in the NRC letter dated November 6, 1986 (E.
G.
Adensam, NRC to C.
A.
McNeill, Jr.,
PSE&G) regarding the Single Loop Operation (SLO)
Technical Specification for Hope Creek Generating Station (HCGS).
In addition, Attachment 2 of this transmittal supersedes of the PSE&G to NRC letter dated May 30, 1986 (C. A.
McNeill, Jr. to E. Adensam).
The remaining information provided in the May 30, 1986 letter is accurate and correct, including the determination that the requested changes to the HCGS Technical Specifications do not involve a significant hazards consideration.
Your expedited review and resolution would be most appreciated.
Should you have any questions or comments on this transmittal, do not hesitate to contact us.
Sincerely, 8612300275 B61224 a
DR ADOCK 050 4
l L
{
Attachments (2)
Director of Nuclear 2-12/24/86 Reactor Regulation i
+
1 C
Mr.
D.
H. Wagner USNRC Licensing Project Manager Mr.
R. W. Borcha rd t USNRC Senior Resident Inspector Mr. Gerald Nichols, Assistant Director Bureau of Radiation Protection Department of Environmental Protection 4
380 Scotch Road Trenton, NJ 08628 Honorable Charles M. Oberly, III j
Attorney General of the State of Delaware 820 North French Street Wilmington, DE 19801
}
4 i
l l
i I.
ATTACHMENT 1 Reprinted below are eight requests for additional information (RAIs) contained in a letter from E. G. Adensam (NRC) to C. A.
McNeill, Jr. (PSE&G) dated November 6, 1986.
These RAIs make reference to-pages, tables and attachments which are references to the letter from C. A. McNeill, Jr. to E. Adensam dated May 30, 1986 (the original SLO submittal).
Each. response providr.d below completely addresses the RAIs and when necessary, identifies appropriate revisions to the Technical Specifications contained in Attachment 2.
1.
In Attachment 1, Table 3.3.6-2, Item 1.C.,
what is the justification for changing rod block monitor downscale trip setpoint to > 4 percent of rated thermal power?
This change was an editorial error made in the development of the May 30, 1986 original single loop operation (SLO) submittal.
As such, the rod block monitor downscale trip setpoint will remain at 5% of rated thermal power as shown in the attached Technical Specifications (see Attachment 2).
2.
Is there a recirculation pump discharge bypass valve in each recirculation loop?
A review of P&ID M-43-1, reproduced in the FSAR as Figure 5.4-2, reveals that no bypass line or bypass valve exists on the discharge line of the reactor recirculation pump.
The A pump's discharge valve, HV-F031A (HV-F031B for B pump), is the only valve between the reactor recirculation pump and the jet pumps.
These valves are provided with 3/4-inch vent and drain lines which are not bypass lines.
3.
Describe the operating position of the recirculation pump discharge valve in the idle loop.
Describe the surveillance performed on its position.
Confirm that the position of the valve was considered in the General Electric analysis.
The General Electric SLO transient analysis was performed assuming (i) the recirculuation pump discharge valve in the idle loop was open, and (ii) the operating loop was operating at maximum power and flow.
The minimum critical power ratio (MCPR) limits calculated from the transient analysis are based on this maximum power / maximum flow point which bounds all operation within the defined operating domain of Technical Specification 3.4.1.1.
With the core flow within the analysis bounds, any flow within the external inactive loop has no bearing on the MCPR limits calculated for the HCGS core.
The LPCI injection path at HCGS is independent of the recirculation loop - injecting directly into the vessel.
For this reason, the Technical Specification surveillance for this valve was eliminated for HCGS.
However; for example, a
O Technical Specification surveillance requirement for the recirculation pump discharge valve is provided at Susquehanna since LPCI injects into the recirculation loop and the ability to demonstrate the operability of these valves is necessary.
4.
In Attachment 1, page B3/4 4-1, the bases discuss recirculation pump speed mismatch.
The Technical Specification discuss flow mismatch.
Bases should be changed to be cor:sistent with proposed technical specification.
The confusion over recirculation pump speed and flow is historical and has existed because of the differences in application and usage of a Technical Specification and a Technical Specification basis.
Regardless, there is no distinction between recirculation pump flow and speed for a BWR/4.
However, in order to clarify the SLO issue for HCGS, the originally submitted Technical Specification basis B3/4.4.1 has been revised to be consistent with Technical Specification 3/4.4.1.3, i.e.,
use of recirculation loop flow only (see Attachment 2).
5.
Confirm that the proposed Technical Specification changes are consistent with General Electric SIL-380, Revision 1.
PSE&G can confirm that the operating recommendations in Service Information Letter (SIL) No. 360, Revision 1 are adequately reflected in the SLO submittal.
6.
In Attachment 1, Table 2.2.1-1, the thw = 8 percent, unless already known, should be indicated as "to determined at a later date."
PSE&G concurs with this recommendation and has appropriately revised Technical Specification Table 2.2.1-1 (see Attachment 2).
7.
In Attachment 2, Table 15.C.3-4:
a)
The peak neutron flux appears to be low (compared with other facilities).
Explain.
b)
What is the " Margin to Safety Limit" in two-loop operation, for the FWCF and LRBPG transients?
a)
Although both Susquehanna and HCGS are BWR/4s, there are some fundamental differences between the two plants; including, (1) a lower rated feedwater temperature for Susquehanna, and (2) a more conservative rod pattern to account for a hard bottom exposure for Susquehanna.
The combination of these two differences means that the end-of-cycle axial power profile for the two plants are
'diffor:nt, with HCGS hnving a powsr peak at nods 5 cnd Susquehanna having a power peak at node 9.
Ths imp;ct of this power shape difference is a less effective scram reactivity for Susquehanna than HCGS because the control rods will reach the high power nodes later for Susquehanna than HCGS.
On the basis of these considerations, the peak neutron flux values will be greater in the Susquehanna core than in the Hope Creek core.
Each core is monitored for target core burnup throughout each plant cycle which will demonstrate any potential deviation from the projected burn path.
Thus, each plant has a continuous check for the validity of the licensing basis assumptions used for the licensing analysis.
With regard to the Peach Bottom Turbine Trip Tests, these tests were conducted specifically for computer model qualification.
These tests were conducted with the j
direct scram or turbine stop valve position bypassed such j
that trip on high flux was obtained.
This departure from i
the normal reactor condition was required to obtain a sufficiently large flux response to allow model-test comparison.
The initial conditions used for the Peach Bottom-2 tests do not correspond to the initial conditions used in the HCGS SLO analysis.
While the initial power levels are approximately the same (8%
different), the initial core flows are not the same (101 Mlbs/hr for Peach Bottom-2 vs. 60Mlb/hr for HCGS.
The increased core flow at Peach Bottom Station results in a decrease in voids which leads to a less bottom peaked power distribution.
Thus, the impact of the increased core flow and the resultant change in power shape is a less effective scram reactivity for the Peach Bottom tests.
Hence, direct comparison with Hope Creek Generating Station is inappropriate.
b)
For two loop operation, the HCGS operating limit minimum critical power ratio (OLMCPR) is given in HCGS Technical Specification Figure 3.2.3-1.
This Technical Specification was revised on December 9, 1986 (Facility Operating License NPF-57, Amendment 1) and as such, the OLMCPR is now a function of the relative scram times and the operability or inoperability of the End-of-Cycle Recirculation Pump Trip (EOC-RPT) function.
Assuming EOC-RPT operable, and ODYN A scram times (T = 1 in Technical Specification Figure 3.2.3-1), the OLMCPR is l.20, and the delta CPR values for two loop operation for Feedwater Controller Failure-Maximum Demand (FWCF) and Generator Load Rejection with Bypass Failure (LRBPF) are 0.09 and 0.07, respectively (HCGS FSAR Table 15.0-1).
Hence, the margins to the Safety Limit CPR of 1.06 are 0.0 for FWCF and 0.02 for LRBPF, allowing for the ODYN Option A adjustments (SER for NEDO-24154 and NEDE-24154-P).
If credit is taken for the modeling bias resulting from the application of the ODYN Option A adjustment, the corresponding margins are 0.05 for the FWCF and 0.07 for the LRBPF.
n Assuming EOC-RPT inoperable and ODYN Option A scram times, the OLMCPR is 1.28.
The limiting delta CPR for the analysis is 0.17.
Allowing for the ODYN Option A adjus tments, the margin to the Safety Limit CPR of 1.06 is 0.0.
If credit is taken again for the modeling bias, the margin is 0.05.
Assuming ODYN Option B scram times (7' = 0 in Technical Specification Figure 3.2.3-1), the current safety analysis is limited by the non-overpressurization transients of Rod Withdrawal Error and Loss Of Feedwater Heating, regardless of the operability of the EOC-RPT function.
8.
Was pump seizure of the pump in the operating recirculation loop considered in the single-loop operation analysis?
If not, discuss this item.
As a result of telecon of December 22, 1986 between the NRC, PSE&G and General Electric (GE), this question was further refined.
The NRC stated that since the LOCA and pump seizure events have different licensing criteria, the LOCA analysis cannot be used to bound the pump seizure event.
Specifically, the NRC stated that the design basis accident (DBA) for a recirculation pump shaft seizure during SLO must be demonstrated to be within a small fraction of 10CFR100 dose rate limits.
A postulated seizure of a recirculation pump is considered a limiting fault event that can satisfy the acceptance criteria of an event of greater probability, such as an infrequent incident classification which must meet the 10CFR100 "small fraction" release criteria.
GE has completed single loop analyses for 23 domestic BWRs and 4 overseas BWRs thus establishing a database of SLO information and analysis techniques which can be applied to new SLO plants, such as HCGS.
Initially recirculation pump seizure analysis were included for all plants requesting the SLO option.
However, the results of the SLO anlayses have always demonstrated that SLO was a non limiting event and boiling transition was not experienced during a recirculation pump seizure-in fact, there was significant margin to the safety limit MCPR.
This is a conservative approach since an infrequent event or a limiting fault do not have to satisfy the safety limit MCPR licensing requirements.
Because the safety limit MCPR is not exceeded, fuel failure is not expected; therefore, the 10CFR100 "small fraction" limits are satisfied for pump seizure during SLO.
Since SLO is a non limiting event and since significant margin to the safety limit MCPR has been previously demonstrated for other BWRs, a plant specific analysis for HCGS is not necessary.
As an example of an analysis for a similar facility, see the SER for Amendment No. 16 to Facility Operating License NPF-29, Grand Gulf Nuclear Station, Unit 1.
P-'
9 9
ATTACHMENT 2
r S 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10%
of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
THERMAL POWER, High Pressure and High Flow IY 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 withfthe reactor vessel steam dome pressure greater than 785 psig x
and core flow greater than 10% of rated flow.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
!WERT ACTION:
2.
With MCPR less than 1.06 @ he reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE I
2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
l APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3 and 4.
l ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
HOPE CREEK 2-1 l
TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS r
ALLOWABLE Q
FUNCTIONAL UNIT TRIP SETPOINT VALUES A
7 1.
Intermediate Range Monitor, Neutron Flux-High
< 120/125 divisions i 122/125 divisions of full scale of full scale 2.
Average Power Range Monitor:
a.
Neutron Flux-Upscale, Setdown
-< 15% of RATED THERMAL POWER
-< 20% of RATED THERMAL POWER f
- 0-@~A*M I b.
Flow Biased Simulated Thernal Power-Upscale l
\\
0.55 m;1%,Pwith with
- 1) Flow Biased Fl : a maximum of 7
a maximum of
==
"I
- 2) High Flow Clamped
< 113.5% of RATED
$ 115.5% of RATED THERMAL POWER THERMAL POWER c.
Fixed Neutron Flux-Upscale
-< 118% of RATED THERMAL POWER
< 120% of RATED THERMAL POWER m
S d.
Inoperative NA NA e.
Downscale
-> 4% of RATED
> 3% of RATED THERMAL POWER THERMAL POWER 3.
Reactor Vessel Steam Dome Pressure - High
< 1037 psig 5 1057 psig 4.
Reactor Vessel Water Level - Low, Level 3
-> 12.5 inches above instrument
> 11.0 inches above zero*
instrument zero 5.
Main Steam Line Isolation Valve - Closure
$ 8% closed 5 12% closed 6.
Main Steam Line Radiation - High, High
$ 3.0 x full power background 5 3.6 x full power background
- See Bases Figure B 3/4 3-1.
INGERr 3
(s...
2.1 SAFETY LIMITS BASES
2.0 INTRODUCTION
The fuel cladding, reactor pressure ves'sel and primary system piping are the principal barriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.
Because fuel damage is not directly g
observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.066. MCPR areater thar,1.06Arepresents a con-q servative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or :se related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signifi-cantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore', the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a signi-ficant departure from the condition intended by design for planned operation.
l 2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow.
Therefore, the fuel cladding integrity Safety Limit is established by other means.
This is done by establishing a limiting condition on core THERMAL POWER with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flowswillalwagsbegreaterthan4.5 psi.
Analyses show that with a bundle flow of 28 x 10 lbs/hr, bundle pressure drop is nearly independent of bundle l
l power and has a value of 3.5 psi.
Thus, the bundle flow with a 4.5 psi driving l
head will be greater than 28 x 103 lbs/hr.
Full scale ATLAS test data taken l
at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-l cal power at this flow is approximately 3.35 MWt. With the design peaking I
factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
l HOPE CREEK B 2-1
Bases Table B2.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT
- Standard Deviation Quantity
(% of Point)
Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5
! 24RWtr Core Inlet Temperature 0.2 5
l
- l Cere Tetel flew
- 2. 51_.__
Channel Flow Area 3.0 i
Friction Eactor Multiplier 10.0 l
Channel Friction Factor Multiplier 5.0 6NSrJer
\\eIT!? R:edi,;;
C. 2 l.
R Factor 1.5 Critical Power 3.6 l
l i
- The uncertainty analysi's used to establish the core wide Safety Limit MCPR is based on-the assumption of quadrant power symmetry for the reactor core.,p H4SEgr
- F l
HOPE CREEK B 2-3
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits iggg gp shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5.f g
8 APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, or 3.2.1-5, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be egual to or less than the limits determined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 and 3.2.1-5:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l
b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and I
l c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
d.
The provisions of Specification 4.0.4 are not applicable.
HOPE CREEK 3/4 2-1
POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION i
3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:
TRIP SETPOINT ALLOWABLE VALUE' o,g[h.@ a S 5 @ % 51%)T S$7
+ 54%)T S
t RB 5 2 4+ 42%)T S
5
+ 45%)T RB where:
S and S are in percent of RATED THERMAL POWER,
{
W=LoohBrecirculation flow as a percentage of the loop recirculation flow which produces.a rated core flow of 100 million Ibs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD).
T is applied only if less than or equal to 1.0.
APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or j
equal to 25% of RATED THERMAL POWER.
ACTION:
[
With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for
(
S or S as above determined, initiate corrective action within 15 minutes andadhs,tSand/ ors to be consistent with the Trip Setpoint values
- within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or red,ce THERMAL POWER to less than 25% of RATED THERMAL u
POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and the CMFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.
Jnitially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with CMFLPD greater than or equal to FRTP.
d.
The provisions of Specification 4.0.4 are not applicable.
Y "With CMFLPD greater than the FRTPt.; te ^^* ef "ATC^ T:lC""AL ^^WC^.)rather than adjusting the APRM setpoints, the# APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times CMFLPD provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice gg of adjustment is posted on the reactor control panel.
i 3
' HOPE CREEK 3/4 2-7 N
i l
TABLE 3.3.6-2
~
CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS k
TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE
[
1.
ROD BLOCK MONITOR
\\
=
a.
Upscale
< '^ EE S t 40%
< " EE " + 43%
I h
b.
Inoperative HA T
HA T
Q o, 4 w. g g c.
> 3% of RATED THERMAL POWER Downscate
> 5% of RATED THERMAL POWER
{
2.
\\
\\
j a.
Flow Biased Neutron Flux -
\\
\\
Upscale
< =. Z 2 + 42%*
< = Z M + 45%*
l b.
Inoperative HA T
HA T
i c.
Downscale
> 4% of RATED THERMAL POWER
> 3% of RATED THERMAL POWER d'.
Neutron Flux - Upscale, Startup 312%ofRATED.THERMALPOWER 314%ofRATEDTHERMALPOWER 3.
SOURCE RANGE MONITORS a.
Detector not full in NA NA 5
5 b.
Upscale
< 1.0 x 10 cps
< 1.6 x 10 cps i
c.
Inoperative HA NA d.
Downscale
> 3 cps **
> 1.8 cps 4.
INTERMEDIATE RANGE MONITORS mg a.
Detector not full in NA NA b.
Upscale 5 108/125 divisions of
< 110/125 divisions of full scale Tull scale c.
Inop'erative NA NA d.
Downscale
> 5/125 divisions of
> 3/125 divisions of l
Tull scale Tull scale l
S.
a.
Water Level-High (Float Switch) 109'1" (North Volume) 109'3" (North Volume) i 108'11.5" (South Volume) 109'1.5" (South Volume)
I 6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.
Upscale
< 108% of rated flow
< 111% of rated flow b.
Inoperative NA NA c.
Comparator 5 10% flow deviation 5 11% flow deviation l
7.
REACTOR MODE SWITCH SHUTDOWN POSITION NA NA WSEEtr y
9 "The-c;;r :: :': :r :::n : nsat
- r. rod block function is varied as a furrMou of recirculation loop flow u
(W)) The trip settina ofA E53 function must be maintained in accordance with Specification 3.2.2.
gy
- May be reduced to 0.7 cps provided the signal-to-noise ratio is > 2 i*'
Iroe 71s w m p_ ceae LeAmm om.y.
f
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:
a.
Total core flow greater than or equal to 45% of rated core flow, or b.
THERMAL POWER less than or equal to the limit specified in Figure 3.4.1.1-1.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.
ACTION:
a.
With one reactor coolant system recirculation loop not in operationt.
gg
' z::::::!y initt t: ::tten :: recur: ~ 5 S"". '. =~in t: :::: th:n er 0';U:1 t0 the l#-it Op Ci# fed #1 ff;;ur 3.'.l.1-1 with h 2 heur and J g
6ftict: ::::;r;; t ph:: th: unit S at 1 ::t "^T SML'700h" withi; i
-12 5 ;r:.
I b.
With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C44 With\\two reactor coolant system recirculation loops in operation and M
c.
total core flow less than 45%dof rated core flow and THERMAL POWER Btr(: RIAE /
greater than the limit specified in Figure 3.4.1.1-1:
THAM (M)%*
P l
1.
Determine the APRM and LPRM** noise levels (Surveillance'?.t.1.1.3):
g
,MIlM a)
At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and b)
Within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.
2.
With the APRM or LPRM** neutron flux noise levels greater than /
! gg g gg three times their established baseline noise levels,V x:::::alv:
initiate corrective action to restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow to greater,than 45% of rated core flow or by reducing THERMAL POWER gg to less than or equal to the lirait specified in Figure 3.4.1.1-1.
t*2
~
- See Special Test Exception 3.10.4.
a
~8$
HOPE CREEK 3/4 4-1
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS M-M P
N
' " 1.1.1
^t le::t once per 12 heur: ver fy that tet:1 cer: '! w i: ;;r^ ter i
L c:r er c';u21 te 45% ef rated cere #!c':? :nd/cr that THER"^.L P'">EP i
!c:: th r th; limit :p :i' icd '- ri;;;r: 2.t.1.1-1.
y i
". t. 1. 1. '" Each pump MG set scoop tube mechanical and electrical stop shall be dem nstrated OPERABLE with overspeed setpoints less than or equal to 105% and 844.l.03 102.5%, respectively, of rated core flow, at least once per 18 months.
3--
F ". t.1.1. 2 Establish a baseline APRM and LPRM** neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) j
)gg,ggy' within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage.
i f
i l
- If not performed within the previous 31 days.
LF 1
I HhPECREEK 3/4 4-2 l
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031V8 % 'M3M0d WWB3H13803 HOPE CREEK 3/4 4-3
REACTOR COOLANT SYSTEM JET PUMPS s.
LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS
- M M
y
- 4 A.I.
- Each of the above required
- jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by determining recirculation loop flow, total core flow and triffuser-to-lower plenurdifferential pressure for each jet pump and verifying that no two of the following conditions occur when the recirculation pumps are operating in accordance with Specification 3.4.1.3.
[
^
The indicated recirculation loop flow differs by more than 10% from i
the established pump speed-loop flow characteristics.
[
r The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.
3.
- [ The indicat.ed diffuser-to-lower plenum differential pressure of any l
individual jet pump differs from the established patterns by more than 10%.
MieRT y
(?
STARruP F-oLLMIMS AMY REFUGLjuG L
OMTAGE AND IN ORd R TD eBrA+J SMGLG LDCY OR "tWO L60p c%2AT1oM BASEL.lAle DATA,
- During:-= ::: rte: tes; pre r- ; data shall be recorded for the parameters listed to provide a ba, sis for establishing the specified relationships.
Comparisons of the actual data in actordance with the criteria listed shall commence upon conclusion of the' t -tup test pregr:4 L l
l aAssuwe tarA mat.vas HOPE CREEK 3/4 4-4
' LCOP Fla/,
RECIRCULATION LIMITING CONDITION FOR OPERATION
�P SHAltH 3.4.1.3 Recirculatio eee shall be maintained within:
r a.
5% of i :t eth:r d th ::re ' :4 greater than or equal to 70% of INS 656 rated core flow.
Y s
b.
10% of!:::n th;r w:th : r: '1:wiless than 70% of rated core flow.
OPERATIONAL CONDITIONS 1* 7and2*/
INSEEE sq APPLICABILITY:
ACTION:
y flows With the recirculatiok : r ::::dd different by nore than the specified limits, either:
7 w
a.
Restore the recirculatioc: r ::::d::to within the specified limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or 7
FLD4J y
b.
Declare the recirculation loop of the pump with the slower g not l
in operation and take the ACTION required by Specification 3.4.1.1.
l l
SURVEILLANCE REQUIREMENTS AcoP
=
l FtcW 4.4.1.3 Recirculation'. ::: :;:: shall be verified to be within the limits i
HISHAlCM at least once,per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> P
l l
"See Special Test Exception 3.10.4.
W zo l
l l
l HOPE CREEK 3/4 4-5
M REACTIVITY CONTROL SYSTEMS D
-8ASES 3/4.1.3 CONTROL RODS j
The specifications of this section ensure that-(1) the minimum SHUTDOWN
-MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) limit the potential effects of the rod drop accident.
The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued j
operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.
The requirements for the various scram time measurements ensure that any indication of systematic prob.lems with rod drives will be investigated on a timely basis.
_ Damage within the control rod drive mechanism could be a generic problem, therefore with a withdrawn control rod immovable because of excessive friction i
or mechanical interference, operation of the reactor is limited to a time period l
which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are i
consistent with the SHUTDOWN MARGIN requirements.
The number of control rods permitted to be inoperable could be more than 1
the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown gg j
for invejtigation and resolution of the problem.
mgg i
The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less thanLt::ggtfuring the limiting power transient analyzed in Section 15.4 of the FSAR.
This analysis 4
shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the
- required protection and MCPR remains greater than M The occurrence of i,
scram times longer then those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of j
time with a potentially serious problem.
The scram discharge volume is required to be OPERABLE so that it will be available,when needed to accelt discharge water from the control rods during a i
i reactor scram and will isolate the reactor coolant system from the containment when required.
Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies.
This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure.
Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.
f
' HOPE CREEK B 3/4 1-2 i
~ -, -... _,. _. _ _,, - - - - - _. _ - -. -, - - _ - -. _ _. _ _ _ _ _ - _. - - - - - _. - - -. _ - ~,. - - - - - - -
-+
r I
POWER DISTRIBUTION LIMITS s,g,'
BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) b.
Model Change 1.
Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
2.
Incoporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLD0D pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL.
These changes are listed below.
a.
Input Change 1.
Break Areas - The DBA break area was calculated more accurately.
b.
Model Change 1.
Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.
DEE5Er e
21 3/4.2.2 APRM SETPOINTS 1
The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER. The flow biased simulated thermal power-upscale scram setting and the flow biased neutron flux-upscale control rod block trip setpointf must be adjusted to ensure that the MCPR does not become less thaT1L-fg or that > 1%
l plastic strain does not occur in the degraded situation.
The scram setpoints ggp,7y and rod block setpoints are adjusted in accordance with the formula in Specifi-Lunrr cation 3.2'.2 whenever it is known that the existing power distribution would cause the design LHGR to be exceeded at RATED THERMAL POWER.
i i
i HOPE CREEK B 3/4 2-2
Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters:
Core THERMAL POWER....................
3430 Mwt* which corresponds to 105% of rated steam flow Vessel Steam Ottput...................
14.87 x 10s 1bm/hr which corresponds to 10S% of rated steam flow Vessel Steam Dome Pressure.............
1055 psia Design Basis Recirculation Line Break Area for:
a.
Large Breaks 4.1 ft2 b.
Small Breaks 0.09 fts, Fuel Parameters:
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNOLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kw/ft)
FACTOR RATIO Initial Core 8x8 13.4 1.4 1.20 d# dt A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the FSAR.
- This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.
l MW SUL i
i HOPE CREEK B 3/4 2-3
POWER DISTRIBUTION LIMITS BASES W
3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating _ conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR M and an analysis of abnormal operational transients.
For any abnormal operating transient analysis evalua-tion with the initial condition of the reactor being at the steady state l
operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-j.
sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).
The type of transients eva'"ated were loss of flow, increase in pressure and_ power, positive reactivity insertion, and cool ant tempe rature dec rease. [ "- ' " "- '---- "-' ""- '" ' -----' ' ' "
i f SP".
L't.:n :ff:d t; th: 5:f:ty Lf:ft T P" Of 1.05, th: 7:gir:d =fni;;;
M
- r: tin; 'f=ft T P" Of S;;;fff::ti:n 3.2.3 f: d t:in;d.
l The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-3 that are input to a GE-core dynamic behavior transient computer progr g The code used to evaluate pressurization events is described in NED0-24154(2) and the program L: sed in non pressurization events is described in NEDO-10802 The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with tgsingle channel transient thermal hydraulic TASC code described in NEDE-25149 The principal result of this evaluation is the reduction in MCPR caused by the transient.
t The purpose of the K factor of Figure 3.2.3-2 is to define operating limitsatotherthanratebcoreflowconditions.
At less than 100% of rated factor.
The K flowtherequiredMCPRistheproductoftheMCPRandtheK(edduringaflow factors assure that the Safety Limit MCPR will not be viola increase transient resulting from a motor generator speed control failure.
The K factors may be applied to both (nanual and automatic flow control modes.
7 The K, factors values shown ir$ Figure 3.2.3-2 were developed generically The K factors were and are ap#11 cable to all.BWR/2, BWR/3 and BWR/4 reactors.
l derivedusingtheflowcontrollinecorrespondingtoRATEDTHERMA[POWERat i
l rated core, flow.
For the manual flow control mode, the K factors were calculated such that 7
for the maximum flow rate, as limited by the pump scoop tube set point and the l
corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows.
The ratio of the MCPR calculated at a given point of core flow, divided j
by the operating limit MCPR, determines the K.f i
HOPE CREEK B 3/4 2-4 l
I
d 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM i
n___.
4. m4 6 ___...... ____,__i...
___4....i.
4._
i_.u____.u.
4.
r gg pr:hibi5b[bbki EN E0ElbE55EN'ef +5[hE-ESbbbEb Ef 5bb ECb5 d:rEE[b55'5:bh S'
23 f i:;:r:ti:n 5:: 5::: ;:rf::::d. :::12:ted :nd det:--in:d t: 5: :::::t91:.
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.
Jet pump failure can be detected by monitoring jet pump performance on a A. COP' prescribed schedule for significant degradation.
N y
s
'M*
Recirculationwump.-opeeg mismatch limits are in compliance with the ECCS 2M T LOCA analysis design criterias The limits will ensure an adequate core flow g,
coastdown from eitler recirculation loop following a LOCA. /
25 In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop.
The loop temperature must also be within M Rgtr 50*F of the reactor pressure vessel coolant temperature to 3revent thermal y
shock to the recirculation pump and recirculation nozzles.% 5tn:: tt: :::::nt in 1 the bette ef the v :::1 i: et a !:tfer t - tr:ture th:n th: 00 !:nt 4a the upper re;;iere f the cere, under tre:: er the ve::01 :: uld re: ult 4' the t ; rature diff r:n : w: crc ter than li5'F l
The objective of GE BWR plant and fuel design is to provide stable operation with margin over the normal operating oomain.
However, at the high power / low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g.,
rod pattern, power shape).
To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region.
Stability tests at operating BWRs were reviewed to determine a generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed.
A conservation decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to i
plant variability of decay ratio with core and fuel designs.
This generic region f
has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.
Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels.
In this case the degree of conservatism can be reduced since plant to plant variability would be eliminated.
In this case, adequate. margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.
HOPE CREEK B 3/4 4-1
INSERT 1 two recirculation loop operation and shall not be less than 1.07 with single recirculation loop operation, in both cases with INSERT 2 with two recirculation loop operation or less than 1.07 with single recirculation loop operation and in both cases with INSERT 3 The Average Power Range Monitor Scram function varies as a
~
function of recirculation loop drive flow (w).
es w is defined as the difference in indicated drive flow (in percent of drive flow which produces rated core flow) between two loop and single loop operation at the same core flow. bsw = 0 for two recirculation loop operation.
tsw = "To be determined at a later date" for single recirculation loop operation.
INSERT 4 for two recirculation loop operation and 1.07 for single recirculation loop operation.
INSERT 5 Core Total Flow Two Recirculation Loop Operation 2.5 Single Recirculation Loop Operation 6.0 INSERT 6 TIP Readings Two Recirculation Loop Operation 6.3 Single Recirculation Loop Operation 6.8 INSERT 7 The values herein apply to both two recirculation loop operation and single recirculation loop operation, except as noted.
INSERT 8 The limits of Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 and 3.2.1-5 shall be reduced to a value of 0.86 times the two recirculation loop operation limit when in a single recirculation loop operation.
INSERT 9 and liw which is defined as the difference in indicated drive flow (in percent of drive flow which produces rated cored flow) between two loop and single loop operation at the same core flow.
INSERT 10 The Average Power Range Monitor Rod Block INSERT 11 1.
Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
a)
Place the recirculation flow control system in the Local Manual mode, and b)
Reduce THERMAL POWER to f 70% of RATED THERMAL POWER, and c)
Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2, and d)
Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0.86 times the two recirculation loop limit per Specification 3.2.1, and e)
Reduce the Average Power Range Monitor (APRM) Scram and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specifications 2.2.1, 3.2.2 and 3.3.6, and f)
Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and g)
Perform surveillance requirement 4.4.1.1.2 if THERMAL POWER is < 30% *** of RATED THERMAL POWER or the recircula' tion loop flow in the operating loop is 1 50%
- of rated loop flow.
2.
The provisions of Specification 3.0.4 are not applicable.
3.
Otherwise be in at least HOT SHUTDONN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
INSERT 12 d.-
With one or two reactor coolant system recirculation loops in operation and-total core flow less than or equal to 39%# and THERMAL POWER greater than the limit specified in Figure 3.4.1.1-1,Jwithin 15 minutes initiate corrective action to l
reduce THERMAL POWER to less than or equal to the limit l
specified in Figure 3.4.1.1-1 or increase core flow to i
greater than 39%# within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
INSERT 13 f
- Initial' values.- Final values to be determined during Startup Testing based upon the threshold THERMAL POWER'and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.
Initial values.
Final values to be determined during Startup Testing (core flow with both recirculation pumps at a minimum
-pump speed).
INSERT-14 4.4.1.1.1 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:
a.
Reactor THERMAL POWER is f 70% of RATED THERMAL
-POWER, and b.
The recirculation flow control system is in the Local l
Manual mode, and c.
The speed of the operating recirculation pump is less than or equal to 90% of rated pump speed, and I
d.
Core flow is greater than 39%# when THERMAL POWER is l
greater than the limit specified in Figure 3.4.1.1-1.
4.4.1.1.2 With one reactor coolant system recirculation loop not in' operation, within no more than 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is < 30%## of RATED THERMAL POWER or the recirculation loop flow in t5e operating recirculation loop is 1 50%## of rated loop flows a.
< 145'F between reactor vessel steam space coolant and Hottom head drain line coolant, and b.
< 50*F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and c.
< 50*F between the reactor coolant within the loop not in operation and the operating loop.
The differential temperature requirements of Specifications 4.4.1.1.2b and 4.4.1.1.2c do not apply when the loop not in operation is isolated from the reactor pressure vessel.
i i
I
a e
INSERT 15 Initial values.
Final values to be determined during Startup Testing (core flow with both recirculation pumps at minimum pump speed).
Initial values.
Final values to be determined during Startup Testing based upon threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.
INSERT 16 4.4.1.2 All jet pumps shall be demonstrated OPERABLE as follows:
INSERT 17 b.
During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that no two of the following conditions occur 1.
The indicated recirculation loop flow in the operating loop differs by more than 10% from the established
- pump speed-loop flow characteristics.
2.
The indicated total core flow differs by more than 10%
from the established total core flow value derived from single recirculation loop flow measurements.
3.
The indicated difference-to-lower plenum differential pressure of any individual jet pump differs from established
- single recirculation loop patterns by more than 10%.
The provisions of Specification 4.0.4 are not applicable c.
provided that this surveillance is performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 25% of RATED THERMAL POWER.
INSERT 18 rated core flow with effective core flow **
INSERT 19 during two recirculation loop operation.
INSERT 20 Effective core flow shall be the core flow that would result if both recirculation loop flows were assumed to be at the smaller value of the two loop flows.
1 INSERT 21 For plant operation with single recirculation loop, the MAPLHGR limits of Figures 2.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 and 3.2.1-5 are multiplied by 0.86.
The constant factor 0.86 is l
derived from LOCA analysis initiated from single loop operation to account for earlier transition at the limiting fuel node compared to the standard LOCA evaluations.
INSERT 22 For single recirculation loop operation, loss of nucleate boiling is assumed at 0.1 seconds after LOCA regardless of initial MCPR.
INSERT 23 The impact of single recirculation loop operation upon plant safety is assessed and shows that single loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specifiction 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively.
MAPLHGR limits are decreased by the factor given in Specification 3.2.1, and MCPR operating limits are adjusted per Section 3/4.2.3.
Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possi-bility of excessive core intervals vibration.
The surveillance on differential temperatures below 30%* THERMAL POWER or 50%*
rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculating pump and vessel bottom head during the extended operation of the single recirculation loop mode.
Initial values.
Final values will be determined during Startup Testing based upon the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottomhead, preventing saturation.
INSERT 24 for two recirculation loop operation.
INSERT 25 In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.
INSERT 26 Sudden equalization of a temperature difference > 145'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.