ML20214K634
| ML20214K634 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/15/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214K632 | List: |
| References | |
| TAC-61083, TAC-61357, NUDOCS 8608210364 | |
| Download: ML20214K634 (21) | |
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'g UNITED STATES
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g NUCLEAR REGULATORY COMMISSION D-j WASHINGTON, D. C. 20655
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.16 TO FACILITY OPERATING LICENSE NPF-29 GRAND GULF NUCLEAR STATION, UNIT 1 MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION i
DOCKET NO. 50-416
1.0 INTRODUCTION
By letter dated May 2, 1986, (reference 1) the Mississippi Power and Light Company (MP&L) requested changes to the Grand Gulf Nuclear Station (GGNS) Unit 1 Technical Specifications to permit o'peration in the maximum extended operating domain (MEOD) with (a) up to a 50' F reduction in feedwater temperature and (b) elimination of average power range monitor (APRM) setdown. These proposed changes involve, among other factors, the development of new power and flow dependent relations for maximum average planar linear heat generation rate (MAPLHGR) and minimum critical power ratio (MCPR). A General Electric Company (GE) analysis of the consequences of operation in the MEOD (reference 2) was included in the submittal to justify the proposed changes.
The ME00 includes expansion of the normal power / flow map into two new regions.
One region, which involves operation at rated power at lower than rated core flow rates, is called the extended load line region (ELLR). The other region, which involves operation at core flows at up to 105% of rated flow is called j
the increased core flow region (ICFR). Operation in the ELLR and ICFR permits
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greater operational flexibility and an improved unit capacity factor.
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Reduced feedwater temperatures can arise from the inoperability or degraded performance of individual feedwater heaters or string (s) of feedwater heaters ge210364e60012 p
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2 or by deliberate reduction of feedwater heating. The operation of GGNS with reduction in feedwater temperature while in the normal power / flow regime was evaluated by MP&L in accordance with 10 CFR 50.59. On the basis of th,is evaluation MP&L concluded that operation with up to a 50* F reduction in feedwater temperature (e.g., rated power feedwater temperature reduction from 420* F to 370' F) would not affect any safety limit and would not increase the consequences of any postulated accident. The conclusions of this evaluation and expansion of operating conditions to include this region were reported in reference 3.
A brief discussion of the consequences of this reduction in feedwater temperature is also given in Section 15.1.7 "Feedwater Heaters Out of Service" which was included in the December 1985 update of the GGNS FSAR.
The average power range monitor (APRM) setdown requirement in the current GGNS Technical Specifications requires that the flow-biased APRM trips be reduced (setdown) when the core maximum total peaking factor exceeds the design total peaking factor. This requirement was associated with a now obsolete Hench-Levy Minimum Critical Heat Flux Ratio criterion. With the elimination of APRM setdown, a revision of the power dependent MCPR limit and development of new flow and power dependent MAPLHGR limits is provided to give fuel protection for power peaking effects at low core flows for those transients terminated by scram. The elimination of the APRM setdown does not affect the results of the i
lossofcoolantaccident(LOCA)calculationssincesetdownwasnotincludedin the LOCA calculations for the current Technical Specifications.
During the review of the MP&L submittal on MEOD, the staff requested additional information concerning Section 15.D.4.1 " Abnormal Operating Transients",
Section 15.D.6 " Loss of Coolant Accidents" and Section 15.D.4.3. " Flow Runout Transients" of reference 2.
This information was provided in reference 4. The analysis in the subnittal also used the results of a generic analysis for the loss of feedwater heating (LOFWH) transient. This LOFWH analysis had not been reviewed by the staff. At the staff's request, the licensee provided a GGNS 0
3 plant specific analysis of this event (reference 5).
Finally, the staff requested a copy of the GE report on Feedwater Heating Out of Service (reference 6) which had been used by MPL in making the 10 CFR 50.59 evaluation 4
of reference 3.
This report was also referenced in the GE analysis (reference
- 2) that was provided in the MP&L submittal on ME00.
By letters dated March 31, 1986, May 2, 1986, and June 2, 1986 (reference 7),
l MP&L also requested changes to the Technical Specifications to (1) pemit operation with one recirculation loop out of service, and (2) to include the (GE) Service Infomation Letter (SIL) No. 380, Revision 1, recommendations regarding themal-hydraulic stability concerns for single loop operation.
Presently, the GGNS operating license requires a unit to be in hot shutdown immediately if an idle recirculation loop cannot be returned to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The recent resolution of Generic Issues B-19 and B-59 regarding thermal-hydraulic stability has provided a basis to permit operation in the single loop mode with appropriate restrictions relating to stability concerns (references 8 and 9). GE, in SIL No. 380 Revision 1, addressed these concerns by providing the boiling water reactor licensees generic guidance for actions which suppress thermal-hydraulic instability induced neutron flux oscillations.
4 The licensee has proposed Technical Specifications in accordance with the guidance provided by GE in SIL No. 380, Revision 1.
Specifically, the following changes are requested by the licensee:
I (1) Revision of the Technical Specifications for average power range monitor (APRM) flux scram trip and rod block settings, an increase in the safety limit Minimum Critical Power Ratio (MCPR) value, an' a d
revision to the allowable Average Planar Linear Heat Generation Rate (APLHGR) values.
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4 (2)
Incorporation of requirements in the Technical Specifications which should result in the detection and suppression of thermal-hydraulic instability induced neutron flux oscillations if they should, occur.
2.0 EVALUATION 2.1 Operation in MEOD with Reduced Feedwater Temperature and Elimination of APRM Setdown 2.1.1 Operation in the ME0D with Normal Feedwater Temperatures The General Electric Company analysis of reference 2, which was provided by MP&L as justification for the proposed changes in Technical Specifications, describes the results of an evaluation of the safety impact of operation in the MEOD with normal feedwater temperatures. This evaluation included consideration of abnornal operating transients, LOCAs, containment pressures, load impact on vessel internals, flow induced vibration, anticipated transients without scram (ATWS), and fuel mechanical performance.
Abnormal Operational Transients All abnormal operational transients of Chapter 15 of the FSAR were considered for operation in the ME00. A bounding analysis was performed using a standard BWR/6 plant at the end of an equilibrium cycle with a highly enriched GE6 fuel type. The transients investigated were generator load rejection without turbine bypass (LRNBP), feedwater controller failure to maximum demand (FWCF),
cold loop startup (CLDLP) and flow control valve opening (FCVO).
It was concluded that the current power dependent MCPR limit (MCPR ) bounded these p
cases in the ME00. Additional GGNS plant specific calculations were made for i
1 5
c the LRNBP and FWCF transients to demonstrate that the bounding analysis for the standard plant equilibrium cycle is bounding for GGNS. At the staff's request, MP&L also provided a GGNS plant specific analysis of the loss of feedwater heating (LFWH) transient in the ME0D. This analysis indicated,that the LFWH transient is less severe than the FWCF transient.
Finally, it is noted that the rod withdrawal error (RWE) analysis in Chapter 15 of the GGNS FSAR included the ME00 region. Hence, the current Technical Specifications power dependent MCPR limit.is protection against the RWE for operation in the ME00.
From the above considerations, it was concluded that the current power dependent MCPR is adequate for operation in'the ME0D. We find this acceptable.
i The flow dependent MCPR operating limit (MCPR ) in the current Technical f
Specifications was based on slow recirculation flow runout transients.
For operation in the ME0D, this event was reanalyzed with approved methods to account for initial operation at low flows and a higher power rod line of the j
ELLR. Two new MCPR relations were developed for two settings of the core flow f
limiter giving maximum core flows of 102.5 and 107.0 percent of rated flow. We find this acceptable.
Themal-Hydraulic Stability The staff has completed the generic review related to the thermal-hydraulic stability of BWR cores.
In the evaluation report (reference 12), the staff concluded that GE fuel designs, including those fuels loaded in the GGNS core, meet the stability criteria set forth in 10 CFR Part 50, General Design Criteria 10 and 12, provided that the BWR has in place operating procedures and Technical Specifications which are consistent with the recommendations of GE SIL-380 to assure detection and suppression of global and local instabilities. This evaluation considered operation in the ME00 with reduction in feedwater heating. Since the licensee is implementing the SIL-380 recomendations, we conclude that this concern is satisfactorily resolved for GGNS during cycle 1 for operation in the MEOD with reduced feedwater heating.
4
6 LOCA Analysis At the staff's request, the licensee submitted a revised justification of the consequences of a LOCA in the ME00, in reference 4.
The results, obtained with approved methods, indicate that operation in the MEOD would result in less than a 5* F increase in the peak clad temperatures of Chapter 6 of the GGNS FSAR and that the requirements of 10 CFR 50.46 are satisfied. We find this
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acceptable.
Containment Pressure Response A conservative containment analysis for operation in the ME00 with FWHOS I
resulted in a peak drywell pressure 1.3 psi higher than the value of 22.0 psi in Chapter 6 of the FSAR. However, this is still below the design limit of 30 j
psig.
It was. also stated that the peak suppression pool temperatures, chugging loads, condensation oscillations and pool swell bounding loads were all found to be bounded by the rated power analysis in Chapter 6.
We find this acceptable.
Load Impact on Internals and Flow-Induced Vibration j
In reference 2 it was stated that the effects of increased reactor internal pressure differences, acoustic loads, flow-induced loads and fuel bundle lift j
forces were evaluated and that design limits were not exceeded. This evaluation included the effect of FWHOS. With respect to flow-induced vibrations, GGNS Unit I was the prototype BWR/6 251 plant for the testing to demonstrate that the flow-induced vibration response of the reactor internals is acceptable.
In reference 11 the staff concluded that the tests demonstrate that the GGNS internals are adequately designed for flow induced vibration effects. We find this acceptable.
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7 Overpressure Protection The sizing of the main steam safety valves for the ASME overpressure p'rotection analysis is obtained for an MSIV closure event with flux scram. Calculations of this event for operation in the MEOD indicated a peak vessel pressure of 1262 psig, well below the ASME code limit of 1375 psig. We find this acceptable.
2.1.2 Operation in the MEOD with Reduced Feedwater Heating The safety impact of operating GGNS Unit 1 in the normal power / flow region at reduced feedwater temperatures was evaluated previously by MP&L (reference 3).
On the basis of a General Electric Company analysis (reference 6), they concluded that a reduction in rated power feedwater temperature of up to 50' F (downfrom420'Fto370'F)wouldnotaffectanysafetylimitsandwouldnot i
increase the consequences of any postulated accident. Hence they modified the GGNS procedures to expand the operational feedwater temperature band to include this 50' F reduction. This evaluation and action by MP&L was reported to NRR in reference 3 in accordance with 10 CFR 50.59 requirements.-
The present submittal by MP&L requests approval for operation in the MEOD with this same reduction in feedwater temperature.
In the General Electric Company analyses of reference 2 which were provided as justification for this request, it is concluded that for operation in the ME00, a reduction in rated feedwater temperatures from 420' F to 370' F would not result in changes to the current MCPR and MAPLHGR limits. We find this acceptable.
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2.1.3 Elimination of APRM Setdown In the current GGNS Technical Specifications the flow-biased APRM trips are reduced (setdown) when the core maximum total peaking factor exceeds the design
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total peaking factor. The General Electric Company analysis (reference 2) supplied with the MP&L submittal includes results from analyses made to detemine the new initial conditions of fuel thermal limits that would be L
needed to satisfy the pertinent licensing criteria if APRM setdown wer'e eliminated. The new limits should 1) prevent violation of the MCPR safety limit, 2) keep the fuel themal-mechanical performance within the design and i
licensing basis, and 3) keep peak cladding temperature and maximum cladding oxidation within allowable limits. The evaluation included operation in the j
ME00 with reduced feedwater temperature.
It was concluded that current MAPLHGR limits protect against a LOCA even without APRM setdown since the current LOCA i
analyses do not take credit for setdown. The flow dependent MCPR limit is also not affected by elimination of APRM setdown since the design basis flow runout
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event is a slow flow / power increase not teminated by scram. However, j
elimination of APRM setdown does affect the power dependent MCPR limit and the j
MAPLHGR limit. The results of the analysis with approved methods are as follows:
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(1) New power dependent relations for MCPR and MAPLHGR Limits are l
provided which include both high and low flow relations at powers j
below 40% where reactor scram on turbine control valve fast closure l
1s bypassed. The MAPLHGR relation is a factor, MAPFAC, which is p
multiplied by the rated MAPLHGR limit to obtain the power dependent i
MAPLHGR limit.
(2) A new flow dependent MAPLHGR factor, MAPFAC. is provided. This f
l factor was determined from analysis of slow flow runout transients with the requirement that peak transient MAPLHGR values not exceed the fuel design basis values.
We find this acceptable.
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2.2 Changes to Technical Specifications to Permit operation in the Maximum Extended Operating Domain The following changes to the Technical Specifications are proposed to 'ermit p
operation of GGNS Unit 1 in the MEOD with feedwater temperatures reduced up to 50' F and with elimination of APRM setdown:
i (1) Table 2.2.1-1(2). The proposed increase of the flow biased APRM setpoint and allowable values of 16% for two loop operation is made to pemit operation in the ELLR part of the MEOD. The GE analysis of reference 2 shows that operation in the MEOD would not exceed design limits. This is acceptable.
l (2) Specification 3/4.2.1. The proposed change to this specification dealing with MAPLHGR limits results from the proposed elimination of i
APRM setdown in Specification 3/4.2.2. The current specifications provide for reduction in the flow-biased APRM trips when the core maximum total peaking factor exceeds the design total peaking factor.
With the proposed elimination of APRM setdown, this peaking effect is covered by revision to the MAPLHGR limits. The revised limits are j
presented as graphs for both a flow dependent and power dependent MAPLHGR factor in Figures 3.2.1-2 and 3.2.1-3.
As discussed in reference 2, the revised limits provide equal or increased margins to fuel integrity limits relative to those obtained with APRM setdown.
We find the proposed changes acceptable.
(3) Specification 3/3.2.2. The proposed change is to eliminate this specification which involves the APRM setdown. As discussed under Specification 3/4.2.1, this proposed change is acceptable.
(4)
Figures 3.2.3-1 and 3.2.3-2.
The slow recirculation flow runout analysis of reference 2 for the proposed operation in the ME00 1
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results in new flow dependent MCPR limit curves. The new curves, shown in Figure 3.2.3-1, are slightly above the curve in the current Technical Specifications. The new set of power dependent MCPR limits shown in Figure 3.2.3-2 result from elimination of APRM setd'own. The new limits include the'effect of operation at feedwater temperature reductions up to 50* F.
The operating limit MCPR at any power / flow condition is the larger of the new flow and power dependent values.
We find the proposed changes acceptabic.
(5) Table 3.3.1-1.
The proposed change involves Note h of Table 3.3.1-1 which deals with bypassing the turbine stop valve closure and turbine
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control valve fast closure scram when thermal power is less than 40%
of rated thermal power. The high pressure turbine first stage pressure is used to measure thermal power. New setpoints for the first stage pressure are provided-for feedwater temperatures greater than 420* F and between 370* F and 420' F.
The proposed change clarifies the current requirement and incorporates the results of the startup tests on power versus first stage pressure. We find the changes acceptable.
(6) Table 3.3.4.2-1.
This revision to Note b of Table 3.3.4.2-1 is a proposed change which clarifies the current requirement and is based on results of the startup test on themal power versus first stage pressure. This change for the end of cycle recirculation pump trip (E0C-RPT) is identical to that for the turbine stop valve and turbine controlvalvefastclosurescram(seeitem5)andisacceptable.
(7) Table 3.3.6-2.
The proposed changes are made to permit operation in the ME00.
Increase in the APRM flow-biased rod block setpoint is proposed to permit operation in the ELLR. However, the high flow clamp to this setpoint value for rod block is added to maintain the same clamp setpoint of 108% as in the current Technical
F 11 Specifications. These changes provide the same margin between the simulated thermal power monitor scram and rod block setpoints as the current Technical Specifications.
In addition, the recirculation flow-high rod block setpoint is increased from 108% to 111% to decrease unnecessary rod block alarms when operating in the ICFR. We find these changes acceptable.
(8) Administrative changes made to the Technical Specifications include elimination of references to the deleted Specification 3.2.2 in Bases 2.2.1 and 3/4.2.2, Specifications 3/4.2.2, and 3/4.1.1, Tables 4.3.1.1-1, 3.3.4.2-1, 3.3.6-2 and 4.3.6-1, the Index, and Figures 3.2.3-1 and 3.2.3-2.
(9) Bases 2.2.1, 3/4.2.1, 3/4.2.2, 3/4.2.3, Bases Table B 3.2.1-1 and Bases Figure B 3/4.2.3-1. Proposed changes to the Bases were those modifications and additions provided to reflect the changes to the Technical Specifications needed for the proposed operation in the MEOD. We find these changes acceptable.
2.3 Single Loop Operation (SLO) 2.3.1 Accidents (Other Than Loss of Coolant Accidents) and Transients Affected by One Recirculation Loop Out of Service One Pump Seizure Accident A plant specific analysis was performed for this event. The analysis has shown that the event results in a MCPR value significantly above the SLO safety limit MCPR.
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2.3.2 Abnormal Operational Transients The licensee discussed the effects of SLO on the course of operational transients.
Pressurization and cold water increase events, as well as' rod withdrawal error, were addressed.
Flow decrease is covered by the pump seizure
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accident already described..The results of calculations for the limiting event for each category were also presented.
Initial operating conditions were conservatively assumed to be 70.6% of rated power and 54.1% of rated core coolant flow.
a)
Pressurization Events
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The limiting pressurization event is the generator load rejection
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without bypass transient.
For single loop operation, the licensee has calculated that the maximum vessel pressure is 1179 psig and the MCPR is 1.41.
Each of the values satisfies its respective safety limit.
i b)
Cold Water Increase Event i
The limiting cold water increase event is the feedwater controller i
failure to maximum demand transient. The reactor is conservatively j
assumed to be in single loop operation at 70.6% of rated power and 54.1% of rated core coolant flow when failure of the feedwater control j
system instantaneously increases the feedwater flow to the pump runout capacity of 130% of rated flow. The peak pressure is calculated to l
be 1059 psig and the MCPR is 1.34, each satisfying its respective safety i
limit.
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. Rod Witndrawal Error l
c) i The rod withdrawal error at rated power is given in the FSAR for the initial core and in cycle dependent reload supplemental submittals.
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These analyses are performed to demonstrate that, even if the operator ignores all instrument indications and the alarms which 4
could occur during the course of the transient, the roS[biock system '
will stop rod withdrawal at a minimum critical power ratio.w'ich is.
h higher than the fuel cladding integrity safety limit. Correction of $
i the rod block equation for single-loop operation assures that the 5
MCPR safety limit is not violated.
One-pump operation results in backflow through 12 of 24 Jet pumps
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while flow is being supplied to the lower plenum from the $ctive jet
.T' pumps.
Because of this backflow through the inactive jet pumps, l
the present rod-block equation and APRM settings must be modified.
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The licensee has modified the two-pump rod block equation and APRM
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settings that exist in the Technical Specification for one-pump operation and the staff has found them acceptable.
The staff finds that one loop transients and accidents other than LOCA, are l
bounded by the two loop operation analyses and are therefore a6ceptable. Loss,
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I of coolant accidents are discussed in Section 2.3.4 below..
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i 2.3.3 MCPR Uncertainties For single-loop operation, the MCPR fuel cladding integrity safety limit is increased by 0.01 to account for increased uncertainties in the total core coolantflowandTraversingIn-coreProbe(TIP) readings. The limiting trans-
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ients were analyzed to verify that there is more than enough margin during SLO to compensate for this increase in safety limit.
A feedwater controller failure initiating at 70.6% of rated power and 54.1% of rated core coolant flow results in a transient delta critical power ratio (CPR)
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of 0.07. A generator load rejection with bypass failure initiated at the-see initial conditions resulted in a transient delta CPR of 0.002.
Since the initial l
operating limit in SLO is equal to or greater than that at rated power and the s
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v transient delta CPR is less in SLO, there is more margin to the safety limit in SLO than at rated power.
For single loop operation at lower power or at lower core coolant flows, the steady-state operating MCPR limit is established This ensures the 99.9% statistical lim'it require-by the MCPR and MCPR curves.
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ment is always satisfied for any postulated abnormal operational occurrence.
Since the maximum core coolant flow runout during single loop operation is only about 54.1% of rated core coolant flow, the current flow dependent MCPR limits j
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which are generated based on the flow runout up to rated core flow are also ade,quate to protect the flow runout events during single loop operation.
Since the SLO transient analysis is bounded by the two-loop transient analysis, power dependent MCPR curves used for two-loop operation are also applicable for SLO.
The staff finds the licensee's consideration of MCPR uncertainties to be acceptable.
2.3.4 Loss of Coolant Accident (LOCA) 1 The licensee has performed analyses of a spectrum of recirculation suction line breaks for single loop operation conditions. The licensee states that evaluation f'
of these calculations, which are performed according to the procedure outlined in NED0-20556-2, Rev.1, indicates that a multiplier of 0.86 should be applied to the MAPLHGR limits for single loop operation of GGNS. This evaluation methodology has been approved by the staff (reference 10).
The principal LOCA concern associated with single-loop operation is the possibility of the LOCA break occurring in the operating loop, in which case there is no coastdown of an intact loop recirculation pump to sustain jet pump and core coolant flow during the early portion of the system blowdown. An early boiling transition may result from this early loss of coolant flow capability.
To account for this possibility, GE derived a single-loop operation MAPLHGR O
multiplier of 0.86 to be used with calculated two-loop MAPLHGR limits during a
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r 15 single-loop operation. The analys$s which determined this multiplier assumed a near instantaneous boiling transition (0.1 second) even though a longer boiling
' transition time may have been calculated using approved sodels. This assumption is very conseW ative whensapplied to the GE fuel.
The MAPLHGR limits developed for MEOD, FWHOS and the APRM setdown elimination are more conservative than those for which SLO was analyzed. The flow dependent MAPLHGR reduction factor is clamped for SLO flows above 59% of rated core coolant flow in order to limit the factor to it's analyzed value of 0.86 i
for SLO. Similarly, the power-dependent MAPLhGR reduction factor is clamped at the 70% of rated power value for SLO, because SLO is only pennitted up to this power level. At 70% power, power dependent MAPLHGR reduction factor is 0.845 which is conservatively below the factor.of 0.86. The staff finds that the consequences of LOCA for SLO are acceptable'with the proposed reduction in MAPLHGR factors.
2.3.5 Thermal-Hydraulic Stability in Single Loop Operation We have evaluated the licensee's proposed Technical Specification changes to assure that the changes provide adequate detection and suppression of potential thermal-hydraulic instabilities.
GE recently presented the staff with stability test data which demonstrated the occurrence of limit cycle neutron flux esci11ations at natural circulation and
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several percent above the rated control rod line. The oscillations were observ-able on the APRMs and were suppressed with control rod insertion.
It was predicted that limit cycle oscillations would occur at the operating condition tested; i
however, the characteristics of the observed oscillations were different from those previously observed during other stability' tests. Namely, the test data showed that some local power range monitor (LPRM) indications oscillated out of phase with the APRM signal and at amplitude as great as six times the core average.
GE has prepared and released a service information letter, SIL No. 380,.to alert i
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the BWR owners of these new data and to reconnend actions to avoid and control abnormal neutron flux oscillations.
The General Electric recommendations were reviewed by the staff and fo'und to be prudent recommendations which provide adequate detection and suppression of potential thermal-hydraulic instabilities as required by General Design Criteria (GDC) 10 and 12 of 10 CFR Part 50. The staff compared these recom-mendations with the GGNS technical specifications for operation with a recir-culation loop out of service. The staff found that the proposed changes are in conformance with the SIL No. 380, Revision 1, reconnendations and are therefore acceptable.
2.3.6 Jet Pump Surveillance Some general questions have arisen regarding the adequacy of surveillance methods which have been used in some plants to monitor jet pump operability during SLO. These methods are used in accordance with NUREG/CR-3052 to close out problems presented in IE Bulletin 80-07, "BWR Jet Pump Assembly Failure."
Since all the hold down beams at GGNS Unit 1 are an improved, acceptable design, the questions regarding SLO jet pump surveillance adequacy are not applicable.
2.3.7
SUMMARY
ON SINGLE LOOP OPERATION i
The staff concludes for GGNS Unit 1 that with the provisions given below, l
transient and accident bounds will not be exceeded during SLO operation.
1.
Minimum Crilical Power Ratio (MCPR) Safety Limit will be Increased by
-0.01 to 1.07 The MCPR Safety Limit will be increased by 0.01 to account for increased uncertainties in traveling in-core probe (TIP) readings. The licensee has determined that the change conservatively bounds the uncertainties introduced by single loop operation.
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17 2.
Minimum Critical Power Ratio (MCPR) Limiting Condition for Operation (LCO)
The licensee proposed that the operating limit MCPR be established by 'the MCPRp and MCPR curves. This LCO is acceptable.
f 3.
The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Limits will be Reduced by Appropriate Multipliers The licensee proposed to reduce the flow dependent MAPLHGR by a factor of 0.86 and the power dependent MAPLHGR by a factor of 0.84 for Single Loop Operation. These reductions are acceptable.
4 4.
The APRM Scram and Rod Block Setpoints will be Reduced 4
The licensee proposed to modify the two loop APRM Scram, Rod Block and Rod Block Monitor (RBM) setpoints to account for back flow through half the jet pumps. These setpoint equations will be changed in the GGNS TS. The changes are similar to other plant TS changes and are acceptable to the staff.
5.
The Recirculation Control will be in Manual Control The licensee proposed to operate the recirculation system in the manual mode to eliminate the need for control system analyses and to reduce the effects of potential flow instabilities. This change is acceptable.
2.4 Changes to Technical Specifications to allow Operation with One Recirculation Loop out of Service (1) Specification 2.1.2 and Bases 2.0.
The MCPR Safety Limit has been increased by 0.01 to 1.07 for SLO to compensate for the uncertainties introduced by SLO. This change is acceptable.
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APRM flow biased scram function equations have an added term to account for the difference between single and two loop recirculation pump (drive) flow for the same core coolant flow. This adjustment accounts for the difference between actual and indicat'ed coolant flow and preserves the original relation between limits and effective drive flow. These changes are acceptable.
(3) Bases Table B2.1.2-1 The standard deviations for total core flow and TIP readings are increased to 6% and 6.8% respectively to account for uncertainties during SLO. This change is acceptable.
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(4) Specification 3/4.2.1. MAPLHGR will be multiplied by the smaller of either the flow-dependent MAPLHGR factor (MAPFAC ) of Fig'ure 3.2.1-2, or f
the power dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-3.
This is p
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acceptable.
(5) Table 3.3.6-2.
APRM rod block (flow biased) equations have an added term to account for the difference between single and two loop drive flow for the same core flow. This adjustment preserves the original relation between actual and indicated flow and preserves the original relation between limits and effective drive flow. These changes are acceptable.
(6) Specification 3/4.3.10. This specification will assure that neutron flux limit cycle oscillations are detected and suppressed. This new specification is added to implement the guidance of SIL 380 to detect and suppress limit cycle power oscillations in the high power / low flow region of the power-flow map. The LIMITING CONDITION FOR OPERATION (LCO),
APPLICABILITY, ACTION, and SURVEILLANCE REQUIREMENTS in the Technical Specifications are consistent with the recommendations in SIL 380, Rev-1.
This is acceptable.
i
T o
9 i
19 (7) Specification 3/4.4.1. This section is modified to permit operation with either one or two loops in operation. The LCO is expanded to address operation while one recirculation loop is out of service. This change is based on and justified by the GE analysis of SLO and is acceptable.
(8) Specification 4.4.1.2.1.
Jet pump surveillance is only required for the operating loop. This is acceptable as described in section 2.3.6 of this SER.
(9) Specification 3.4.1.3.
The LCO is changed to reflect that recirculation loop flow mismatch is only of concern when both recirculation loops are in operation. This is acceptable.
2.5 REFERENCES
1.
Letter from O. D. Kingsley, Jr., Mississippi Power & Light Company, to H.
Denton, NRC, " Proposed Amendment to the Operating License (PCOL-86-07) -
Maximum Extended Operating Domain", May 2, 1986.
2.
General Electric Company Report "GGNS Maximum Extended Operating Domain Anal.vsis" March, 1985.
3.
Letter from 0. D. Kingsley, Jr., Mississippi Power & Light Company to J.
N. Grace, NRC, Region II, November 15, 1985.
4.
Letter from O. D. Kingsley, Jr., Mississippi Power & Light Company, to H.
Denton, NRC, " Addendum to MEOD Submittal", July 11, 1986.
5.
-Letter from O. D. Kingsley, Mississippi Power & Light Company, to H.
Denton, NRC, " Addendum to ME00 Submittal", June 9, 1986.
,r---
7.
4
(
20 6.
General Electric Company Report "GGNS Feedwater Heater (s) Out of Service Analysis", March, 1986.
7.
Letters from O. D. Kingsley, Jr., Mississippi Power & Light Compa'ny, to H.
Denton,NRC,AECM-86/0092,AECM-86/0129,AECM-86/0160, dated March 31, May 2, June 2,1986 respectively, i
8.
Generic Letter No. 86-02 " Technical Resolution of Generic Issue B-19 Thermal Hydraulic Stability," January 23, 1986.
9.
Generic Letter No. 86-09 " Technical Resolution of Generic Issue B-59 (N-1)
Loop Operation in BWRs and PWRs," March 31, 1986.
10.
Letter from H. N. Berkow (NRC) to J. F. Quirk (GE) dated March 5,1986.
Acceptance for Referencing of Licensing Topical Report NED0-20566-2, Rev-1, " General Electric Analytical Model for Loss of Coolant Accident Analysis in Accordance with 10 CFR 50 Appendix K, Amendment No. 2, One Recirculation Loop-out-of service."
- 11. Memorandum from G. Lainas to W. Butler, " Grand Gulf Unit 1 Reactor Internals Vibration Measurements and Inspection Program", May 28, 1986.
- 12. Letter from L. S. Rubenstein to D. Crutchfield, " Safety Evaluation of GE Topical Report NEDE-24011 (GESTAR) Amendment 8", April 17, 1985.
3.0 ENVIRONMENTAL CONSIDERATION
The amendment involves a change to requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR 20 and changes to the surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no 9
e 21 significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there' has been no public comment on such finding. Accordingly, this amendment meets the eligi-bility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (51 FR 24258) on July 2, 1986, and consulted with the state of Mississippi. No public comments were received, and the state of Mississippi did not have any comments.
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security nor to the health and safety of the public.
Principal Contributors:
C. Graves and G. Thoras, Reactor Systems Branch, DBL L. Kintner, BWR Project Directorate No. 4, DBL Dated:
August 15,1986 w - -,,.---
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