ML19341C321
| ML19341C321 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/10/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19341C320 | List: |
| References | |
| NUDOCS 8103020735 | |
| Download: ML19341C321 (24) | |
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- UNITED STATES NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE'0F NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 31 TO FACILITY OPERATING LICENSE NO. DPR-69 BALTIM0RE GAS & ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NO. 2 DOCKET NO. 50-318
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t-TABLE OF CONTENTS Page 1.0 Introduction 1
2.0 Discussion and Evaluation 2
2.1 Cycle 4 Fuel Design 2
2.1.1 Thermal Performance Analytical Methods 4
2.1.2 Cladding Creep Collapse 5
2.1.3 Fuel Rod Bowing 5
2.1.4 Fuel Assembly Shoulder Gap 5
2.1.5. CEA and Guide Tube Integrity 6
2.2 Nuclear Analyses 7
2.3 Thermal Hydraulic Design 8
2.4 Accident and Transient Review 8
2.4.1 Loss of Coolant Flow 8
2.4.2 Boron Dilution 8
2.5 Loss of Coolant Accident Review 9
2.6 Radiological Consequences of Postulated 9
Accidents-2.7 Reactor Protection System Asymmetric Steam Generator Transient Protection Trip Function 10 2.8 Reactor Coolant System Vent Installation 12 2.9 Reactor Coolant Pump Stud Corrosion 13 2.10 Hydraulic Snubber Common Reservoirs 13 2.11 Reactor Coolant Pump Lube Oil Collection System 14 2.12 Shutdown Cooling R'elief Valve Testing 15 3.0 Technical Specification Changes to be Completed 15 4.0 Physics Testing 18 5.0 Environmental Consideration 18 6.0 Conclusion 19 7.0 References 20
- 1.0 ' introduction By applications dated December 4 and 12,1980 (1,29 ) and supple-
- mental information as listed in the r,eference section, Baltimore Gas and Electric Company (BG8E or the licensee) requested an amendment to Facility Operating License No. DPR-69 for the Calvert Cliffs Nuclear
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Power Plant, Unit No. 2 (CCNPP-2 or the facility). The amendment request consists of:
s Apendix A (Safety) Technical Specifications (TS) changes resulting from the analyses of the Cycle 4 reload fuel; e Continued approval to operate with modified (sleeved) control element assembly (CEA) guide tubes; and e Approval 'of new reactor protection system (RPS) asymmetric steam generator transient protection trip. function (ASGTPTF).
The associated specific TS changes are described in Section 3.0 of the following Safety Evaluation (SE).
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In addition, this SE addresses our evaluation of:
o Reactor coolant system (RCS) vent installation; e Reactor coolant pump (RCP) stud corrosion; e Hydraulic snubber common reservoirs; e RCP lube ' oil collection system; and s Shutdown cooling system relief valve testing.
These evaluations are presented in Sections 2.8 through 2.11, respectively, of this SE.
By application dated January 29, 1981 (30),, BG&E requested four (4) unrelated TS changes. They state that Number 4, having to do with surveillance testing of RV-469, is necessary for the current refueling outage. This request will be addressed in Section 2.12 of this SE.
By letter dated June 26, 1980, BG&E submitted six volumes of new methodology prepared by Combustion Engineering Company (CE) for use in future reloads of the CCNPP units.
In the Reference 3 application for Cycle 5 operation of CCNPP-1, four of these CE reports were used to generate the Cycle 5 limiting conditions for operations (LCOs) and limiting safety system settings (LSSSs). The reports used were Statistical Combinations of Uncertainties (SCU),
Parts 1, 2 and 3 and FIESTA (a one dimensional, two group space-time kinetics code for calculating PWR scram reactivities).
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The staff determined that insufficient time was available to review the four CE reports prior to the scheduled CCNPP-1 startup for Cycle 5 operation.
BG&E was requested to perform a reanalysis, using only staff approved methods, to generate new LCOs and LSSSs.
Such a reanalysis was submitted on November 4, 1980 (4). We have committed to complete our review of SCU Parts 1, 2 and 3 and FIESTA by early April 1981. At that time another SE will be issued.
Since our review of the six volumes of new methodclogy has not been completed, BG&E elected to use the CCNPP-1 Cycle 5 analysis as a " reference cycle" accounting for specific. core differences in the Reference 1 analysis. BG8E concludes that either the reference cycle analysis provided in References
' 3 and 4 envelopes the new conditions or a revised analysis was completed to. show acceptable results. Our approval of the reference cycle is given in Reference 5.
Other information provided to support this reload and related evaluations is as listed in the references (Section 7.0 of this SE).
2.0 Discussion and Evaluation In this evaluation of the Cycle 4 reload with increased fuel assembly enrichment for longer cycle operation, use is made of our generic review of various topical reports.
Some of the topical reports have received formal NRC approval.
In all cases where a topical report nas not received such an approval, the report has been examined, its methods judged to be reasonable, and an appraisal has been made that a complete review will not reveal the methodology to be significantly in error.
On this basis, all topical references are judged to be acceptable for this reload of CCHPP-2 and for operation at the licensed power level of 2700 MWt.
2.1 Cycle 4 Fuel Design The objectives of the fuel system safety review are to provide assurance that (a) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences, (b) fuel system damage is never so severe as to prevent control rod insertior, when it is required, (c) the number of fuel rod failures is not underestimated for postulated accioents, and (d) coolability is always maintained. We have reviewed the information provided in support of Cycle 4 operation of CCNPP-2 to determine if these objectives have been met. Our evaluation is des-cribed below.
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. The CCNPP-2, Cycle 3 reload was performed via the provisions of 10 CFR 50.59.
The original fuel management pattern'was developed to accomradate a Cycle 3 endpoint exposure range of 10,000 to 11,000 mwd /Mtu. The actual core exposure achieved during Cycle 3 was 11,153 mwd /MtU bringing the core-average End of Cycle (E0C) 3 exposure to 22,538 mwd /MtU (2).
The Cycle 4 core will be comprised of 217 fuel assemblies that were manu-factured by CE, the original NSSS vendor. After the reload, the core-average Beginning of Cycle (B0C) 4 exposure will be about 5,625 mwd /MtU, thus making the predicted EOC 4 core-average exposure about 22,979 mwd /Mtu.
!The Cycle 4 core loading inventory is giver in the following table. The 1ead batch burnup in expected to be 36,900 mwd /Mtu.
1 CALVERT CLIFFS, UNIT 2, CYCLE 4 CORE LOADING INVENTORY Initial BOC E0C Assembly Number of Enrichment Burnup Average Burnup Average Designation Assemblies w/o U235 (mwd /MtU)
(mwd /MtU)
D 25 3.03 22,200 36,900 E
48 3.03 10,400 28,100 E/
16 2.73 F
40 3.65 0
17,700 F/
88 3.03 YT7 Total The major changes to the core for Cycle 4 include the removal of 1 Batch B assembly, 68 Batch C assembl.ies, and 59 Batch D assemblies. These assemblies will be replaced with 40 Batch F (high enrichment) and 88 Batch F/ assemblies.
Tne fresh Batch F fuel will be the limiting batch with respect to stored ene gy.
Tc accommodate extended burnup cycles, each of the Batch F/ fuel assemblies will employ 8 burnable poison pins.
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2.1.1 Thermal Performance Analytical Methods The CE densification kinetics expression, along with data on fuel swelling, thermal expansion, fission gas release, fuel relocation, thermal conductiv-ities, cladding creep, and other properties, have been combined in a detailed fuel performance evaluation model called FATES, which is presented in the CE topical report CENPD-139, " Fuel Evaluation Model" (6). This model was used to calculate fuel temperature and stored energy, linear thermal output, and augmentation (power spikes) factors.
In 1976, after the approval of CENPD-139 (7), information was made available to the NRC that lead us to question the validity of fission gas release calculations in the CE model for burnups greater than 20,000 mwd /Mtu.
CE and BG8E were informed of this concern and provided with a method of correcting fission gas release calculations for burnups greater than 20,000 mwd /MtU (8, 9).
In response to our concern, BG&E provided References 10 and 11 showing the addition of the NRC correction method would not adversely affect CCNPP-2 safety analyses for LOCA or other postulated events.
It was also shown that rod internal gas pressure would remain below nominal coolant pressure during the current cycle of operation. We accepted this information as a basis for continued acceptance of the as-submitted safety analysis provided there was "no extension of these burnups or other factors which significantly affect fission gas release, LOCA PCT or fuel rod internal pressure"(12).
With the planned burnup extension for Cycle 4, the previous basis for continued acceptance is no lo'nger valid. However, a recent BG&E submittal provides a partial basis for continued acceptance.
Reference 13 provides results of rod internal pressure analyses using the present CE fuel performance model with the NRC correction for enhanced fission gas release. The results show that the fuel rod internal pressure will not exceed nominal system pres,are for the maximum burnup pro.iected at E0C 4.
We, therefore, conclude that the end-of-life rod pressure criteria has been appropriately considered for this proposed cycle of operation.
With respect to the impact of enhanced fission gas release on LOCA and other transients and accidents, BG8E has not responded to our questions in this area.
In order to assure that the impact of enhanced fission gas release on LOCA and other safety analyses be considered in the future,,
we have reached an agreement with BG&E and CE to provide this new analysis for CCNPP-1 Cycle 6 and all subsequent reload analyses for the Calvert Cliffs units. We find, on this basis, that the Cycle 4 analysis can be accepted for CCNPP-2 operation.
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5-2.1.2 Ciadding Creep Collapse Combustion Engineering has written a computer code that calculates time-to-collapse of Zircaloy cladding in a pressurized water reactor environ-ment.
This code is described in CENPD-187, "CEPAN Method of Analyzing
. Creep Collapse of Oval Cladding" (14). We have reviewed this code and found it acceptable as described in our safety evaluation, which is bound into Reference 14.
For the extended Cycle 4 operation, CE has performed time-to-cla,iding-collapse calculations using CEPAN conservative input values of internal rod pressure, cladding dimensions, cladding temperature, and neutron flux.
The results of this analysis showed that the minimum time-to-collapse is in excess of the E0C 4 batch-average discharge lifetime of the fuel. We, therefore, conclude that the fuel rod cladding in the Calvert Cliffs, Unit No. 2, Cycle 4 core will not collapse and is acceptable in this regard.
2.1. 3 Fuel Rod Bowing Because fuel rod bowing in pressurized water reactors affects neutronic and thermal-hydraulic safety margins, BG&E analyzed the anticipated extent of rod bowing in Cycle 4.
In the analysis, BG&E has used the staff approved interim acceptance methods by which rod bowing analyses can be made (15).
In Reference 15 we have (a) given approval of the burnup-dependent approach to rod bowing, (b) presented a formulation to be used in extrapolating bow magnitudes to.other designs, and (c) described the factor that increases the cold rod bow magnitudes (which are determined from cold-measured gap closures in spent #uel pools) to account for not rod bow magnitudes that occur in-reactor during hot-operating conditions.
BG8E has found that, for a bounding exposure of 37,000 mwd /MtU, the corres-pounding DNBP, penalty (4.4%) is less than the available DNB margin (6.0% as determined by examining the Cycle 4 power distributions of the affected assemblies ).
We, therefore, conclude that there'is no need for a Cycle 4 rod bowing penalty.
2.1.4 Fuel Assembly Shoulder Gap During irradiation, fuel iods and fuel assembly guide tubes undergo axial i
growth at different rates.
If this differential growth progresses to
'the point of consuming all of the available shoulder gap, then mechanical interference will occur between the fuel rod end caps and the fuel assembly structure. To ensure that an adequate design shoulder gap exists for the fuel assemblies that will comprise the Cycle 4 core, shoulder gap calculations on the lead batch (Batch D) fuel assembly were performed with the methods described in CENPD-198, "Zircaloy Growth In-Reactor Dimensional Changes 6
6-in Zircaloy-4 Fuel Assemblies" (23, 24, 25). We hcve rev! awed the CE topical reports and approved them for referencing; however, our approval was limited to an axially averaged fast neutron fluence of 4 E 21 n/cm, which corre-sponds to a maximum assembly of 22,500 mwd /MtU (26). This is an exposure above which CE has not reported data on their core components.
Assurance of the acceptability of CCNPP-2, Batch D and E fuel assemblies at exposures beyond 22,500 mwd /MtU has, therefore, not been rigorously demonstrated. There is, however, comfort in that analysis with this unverified ~model did show sufficient clearance for the lead-exposure Batch D assemblies at a high (95%) confidence level current axial-growth-strain measurements reported from other sources are lending support ta the validity of _using linear growth extrapolations to fluence levels beyond the current CE data base.
Therefore, we conclude that the concern of adequate fuel assembly shoulder gap has been satisfied for Cycle 4 operation.
2.1.5 CEA and Guide Tube Integrity A fretting wear has been observed (for example see References 16, 17, 18, and 19) in irradiated fuel assemblies taken from operating CE reactors.
.These observations revealed an unexpected degradation of guide tubes that were under control element assemblies.
It was subsequently concluded that coolant turbulence was responsible for inducing vibratory motions in the
.l' normally fully withdrawn control rods and, when these vibrating rods were
-in ~ contact with the inner surface' of the guide tubes, a wearing of the guide tube wall takes place.
Significant wear has been found to be limited to t,e relatively soft Zircaloy-4 guide tube because the Inconel-625' cladding on tqe control rods provides a relatively hard wear surface. The extent of tqe observed wear appears to be plant dependent and has, in some cases, extended completely through the guide tube wall.
As an interim fix, BG8E and other licensees of CE designed NSSS installed stainless steel sleeves ir all new and used fuel assembly guide tubes to be used in CEA positions. Our review of the sleeving programs has been documented in previous SEs (for example see the Millstor.3 Unit No. 2 Cycle 3 reload safety evaluation in Reference 20). Our prior'SEs concluded' l
that ouide tube sleeves will perform their function of reducing guide e
tube 1 stresses to acceptably low values in worn assemblies and that sleeves are satisfactory for mitigating further fretting wear in irradiated or fresh fuel assemblies.
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t As additional evidence of BG&E's satisf actory results with guide tube sleeves, recent surveillance results were submitted (21). This submittal described inspections performed on five Unit-1 fuel assemblies that were discharged during the last reload outage.
There was no significant sleeve wear observed.
Therefore, in light of BG&E's demonstrated success of mitigating CEA guide tube wear with the use of sleeves in all fuel assemblies under CEAs, we conclude that the issue of CEA guide tube wear is, at least in the interim basis, resolved for CCNPP-2.
In addition, our Reference 5 SE concluded that the reduced flow test now in CCNPP-1 is acceptable.
BG&E has agreed to a proposed TS chary: that would limit the'Calvert Cliffs units to such modifications to mitiga.te CEA guide tube wear as have received our documented approval. -We find such a TS change is acceptable and resolves this issue at least in a " permanent interim" basis.
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At a later time when an improved method is demonstrated to be efiective in eliminating CEA guide tube wear, BG&E could propose a second TS change, with the supporting analysis,'to implement the final fix of this problem.
While the installation of a sleeve in all guide tubus to be under CEAs has resolved this issue for CCNPP-2, cladding wear could possibly occur at the point where the CEA tip vibrates against the sleeve.
Therefore, during the Cycle 4 refueling outage, BG&E is performing video, profilometry and I
eddy current examinations with an encircling coil on 3 Unit 2 CEAs in order to determine if any cladding wear has occurred.
It is anticipated that the results (to be submitted for NRC review within the customary 30 days af ter returning Unit 2 to power) f rom these exam-inations will be f avorable and will further substantiate the. results from prior inspections made on CEAs during the Unit 2 Cycle 3 outage and the Uni: l' Cycle 5 outage (21).
To date, no inspections have revealed CEA cladding wear rates.that would indicate a possible loss of CEA hermiticity in the near future.
We can, therefore, conclude that for Cycle 4 operation, f retting wear to CEA cladding will remain at acceptably low levels.
It remains uncertain, however, as to whether wear degradation to CEAs could ultimately reduce the CEA design lifetine.
BG&E has agreed to continue CEA examinations to evaluate this problem.
2.2 Nuclear Analyses In our review of the nuclear analyses performed by CE for BG&E, we comparad the results of the CCNPP-2 Cycle 4 with the reference cycle, CCNPP-1 Cycle 5.
We found the critical nuclear design parameters bounded by the identical TS limits used for CCNPP-1, Cycle 5.
Therefore, the Cycle 4 analysis for CCNPP-2 is acceptable.
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. 2.3 Thermal Hydraulic Casign Our review indicates that the thermal hydraulic models and pertinent design parameters used for the CCNPP-2, Cycle 4 analyses are the same as those used for CCNPP-1, Cycle 5 analyses and are, therefore, acceptable as documented in Reference 5.
The proposed TS thennal hydraulic limits are identical for both units.
2.4 Accident and Transient Review Each design basis event (DBE) considered in the safety analysis report has been reviewed by comparing the current and reference cycle key parameters that significantly impact the results of the event. The results of the review were that the key input parameters to all the DBEs for CCNPP-2, Cycle 4 are the same as those for the reference cycle except for the Loss of Coolant Flow (LOCF) event. We find all other accident and transient analyses acceptable.
2.4.1 Loss of Coolant Flow Event TThe LOCF event was reanalyzed to account for the fact that the ficw coastdown curve for Unit 2 is steeper than that of Unit 1.
The same methodology was used in the current and reference cycles. The LOCF is an undercooling transient with the safety criteria limiting the DNBR to a minimum of 1.195 and limiting the RCS pressure to a maximum of 2750 psia.
The results of the LCCF tansient, as reanalyzed fo'r CCNPP-2, Cycle 4 are a minimum DNBR of 1.195 and a neak RCS pressure of 2307 psia.
The proposed Unit 2, Cycle 4 operation, therefore, has enough safety margin that, in association with the reactor protection system (Low Flow Trip at 93% of initial 4-pump flow), will protect the reactor against the postulated LOCF.
We find these results acceptable.
2.4.2 Boron Dilution In the course of our review of the boron dilution event during cold shutdown for CCNPP-1, Cycle 5 operation, BG8E indicated that the operator has no positive means of being alerted to a boron dilution event that may be in progress.
Since the calculated minimum time of 19.7 minutes may expire and the reactor could become critical before the operator is able to identify and terminate the event, it i's the staff's position that the operator should be provided with a positive i
neans to alert him to a boron dilution event.
This would allow him adequate time to terminate the event. BG&E has agreed to provide a means of such notification during the Cycle 4 operation of Unit 2.
o 9-2.5 Loss of Coolant Accider,t Review BG&E states that the ECCS p -formance analyses, to demonstrate compliance with 10 CFR 50.46, bounds bt~ 1 Calvert Cliffs Units No.1 and 2 for Cycle 5 and Cycle 4 operation, respectively. The following table compares the results of this analysis with our requirements. We, therefore, find the LOCA analysis acceptable.
CALVERT CLIFFS LIMITING BREAK (1.0 DES /PD) RESULTS Peak Cladding Peak Local Core Wide Case Temperature (*F) 0xidation (%)
0xidation (%)
Unit 2/
Cycle 4 1987 9.7
<.51 10 CFR 50.46 2200 17 1.0 2.6 Radiological Consequences of Postulated Accidents BG&E has used as a reference cycle the recently approved Cycle 5 for CCNPP-1.
We reviewed the radiological consequences of' accidents for Cycle 5 and found the results to be an acceptably small fraction of 10 CFR 100 limits.
By comparison of CCNPP-2 Cycle 4 with CCHPP-1 Cycle 5, it was determined that the conclusion of acceptability was applicable in the present case.
The 11censee has provided no analysis of burnup dependent phenomena which l
might impact.the number of rods predicted to fail during accidents or the releases from the failed fuel rods. Therefore, our evaluation is valid for traditional target burnups of 33,000 mwd /MtU, batch average discharge.
. 2.7 Steam Generator Transient Protection Trip Function The asymmetric steam generator transient protection trip function (ASGTPTF) designed for both CCNPP units by CE is to provide a reactor trip for those design basis events (A00s) associated with secondary system malfunctions which result in asymmetric primary loop temperature. The most limiting event is the loss of load to one Steam Generator (LL/lSG) caused by a single Main Steam Isolation Valve (MSIV) closure.
The CCNPP RPS presently employs an analog thermal margin trip calculator as part of the Thermal Margin / Low Pressure (TM/LP) trip function. To provide a reactor trip for asymmetric design basis events, pressure in each of the two steam generators will be mo titored and these signals input to the
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thernal margin calculator.
Secondary pressure imbalances between the two generators will be calculated and a corresponding f actor applied in the TM/LP Calculator to generate a trip signal.
Protection against exceeding the DNBR and maximum kW/ft ^?ecified Acceptable Fuel-Design Limits (SAFDLs) during the LL/ISG event is p.esently provided by the Low Steam Generator Level reactor trip in conjunction with sufficient initial margin maintained by the Limiting Conditions for Operation (LCOs).
The ASGTPTF will result in a reactor trip sooner than the Low Steam Generator Level trip and, hence, will produce a smaller margir degradation during this event.
The additional margin gain allows full advantage to be taken of margin recovery programs designed to achieve 18 month fuel cycles fo-future BG&E reload cycles by assuring that the asymmetric transients would not be-limiting A00s for establishing the LCOs.
This new trip function utilizes existing steam generator pressure sensors and TM/LP calculators which are used to generate trips as part of the RPS.
The TM/LP calculators will be modified to include a bistable with an input of tne absolute value of the pressure difference between the two steam generators.
If the channel cifference exceeds a set amount, a bias is input to the TM/LP calculation causing that channel to trip. The additional bias input to the TM/LP calculation is the asymmetric factor signal. The trip signal is preceded by a pre-trip alarm to alert the operator of undesirable (abnormal) operating conditions.
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l The Calvert Cliffs RPS utilizes four vital buses and a two-out-of-four trip logic.
There are four channels of steam generator pressure per steam generator and four TM/LP channels. Each of the four channels is powered from a separate vital ~ bus. Each TM/LP channel receives a pressure signal r
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. from the corresponding (powered from the same bus) pre:sure channel of each steam generator.
Thus electrical independence between channels is mai ntained.
Loss of power to any TM/LP channel will place that channel in the tripped state.
There are no control or indication functions associated with this trip 4
function, nor is there any interaction with non-safety circuits. The licensee has stated that the ASGTPTF instrumentation conforms to the requirements of IEEE Standard 279-1968 and that the addition of this trip function will in no way degrade or adversely affect the operation of the existing RPS.
A channel functional test will be performed monthly for the ASGTPTF channels.
- In addition, a channel check (steam generator pressure) is performed during each shift and the channcis are calibrated ejch refueling outage.
The ASGTPTF is bypassed during startup below 10 '4. of rated thermal power.
When thermal power is >10-h of rated, this bypass is automatically removed.
The conponents being added for the ASGTPTF are of the same type and quality as those being used in the existing RPS. The trip function is designed so that protective action will not be initiated due to normal operation of the generating station.
The selection of a trip setpoint is such that adequate protection is provided when all sensor and processing time delays and inaccuracies are taken into a ccount.
Final determination of an equipment setpoint is based on equipment characteristics, operating environment, NSSS performance and safety analysis.
The nominal setpoint, uncertairities and response time are provided in the following-table.
ASY!NETRIC STEAM GENERATOR TRANSIENT PROTECTION TRIP FUNCTION NOMINAL SETPOINTS Nominal System Accuracy
- 35 psi Analysis Setpoint
+175 psid Nominal Equipment Setpoint
+135 psid Nominal Pretrip Setpoint
+ 100 psid Nominal System Response Time i.9 seconds Based on our ieview of the licensee's submittal, we conclude that the proposed modifications to the CCNPP-2 RPS resulting in the addition of an ASGTPTF trip are acceptable.
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. 2.8 RCS Vent Installation One of the modifications to be made at all PMRs as a result of the Three Mile Island (TMI) accident is the installation of RCS vents. Guidance was provided on this Lessons Learned Item No. 2.1.9 in our letters of September 13 and October 30, 1979.
Additional guidance has been given in our October 31, 1980 letter under Action Plan Item No. II.B.l.
BG8E provided their conceptual design in their letter of January 4,1980.
Since the RCS vents could only be installed during an outage, the licensee elected to install two vent manifolds to vent the domes of the reactor vessel and the pressurizer during the Cycle 4 reload. Because the operational procedures bave not been developed by BG&E and the staff review is not completed, we find it necessary to review only the portions of the vent design dealing with inadvertent operation for the inter.m period until the entire vent review is completed.
BG&E states.that the hardware modifications include the installation of two 3/4" vent manifolds each containing two series isolation valves, one located on the reactor vessel head and the other at the top of the press-urizer. Both manifolds are installed to existing penetrations of the reactor vessel and the pressurizer heads. The manifolds will discharge through 1/2" stainless steel tubing into a common 1/2" line connecting into an existing 10" line immediately upstream of the quench tank. The existing quench tank vent system will be replaced with valving to enable the control room operator to vent this tank. to the waste gas system or the containment as required.
Each vent manifold contains two remotely operated valves in series.
These solenoid operated valves are normally closed and are designed to fail closed.
Each valve is controlled by a handswitch in the control room which is installed with a lock. Power is supplied by emergency buses and each valve has separate. fuses. To ensure that the potential for inadvertent operation is minimized, BG8E has committed to deenergize the valve operating solenoids by opening and tagging a pair of sliding links in the control room. This will be the same method used for Unit 1 (27).
The BG&E analysis shows that the stainless steel line size (1/2") has been sel y ted to ensure that the vent discharge from the reactor vessel, assuming saturated water at 2500 psia, would not exceed the make-up capacity of
'one charging pump (44 gpm). They find this results in no additional LOCA analysis being required for this installation.
We find that since (1) previously existing penetration's of the reactor vessel and pressurizer heads are utilized in this modification; (2) each
.nanifold contains two valves in a serie, arrangement to ensure isolation ability; (3) all valves will be remotely disabled by removing the operating power during plant normal operation; and (4) discharge flow of either manifold would not result in a loss of RCS inventory since it would not exceed the
d
. makeup capacity of one charging pump, this modification is acceptable for return to reactor operation until the entire vent system is completed.
The review of the operating criteria and our other requirements for this system will be completed at a later time. We believe, however, that since this modification is completely installed at CCNPP-2, the licensee should expedite the development and submittal for NRC review of operating procedures and TS..BG&E has agreed to do this.
2.9 Reactor Coolant Pump Studs I&E Information Notice 80-27 dated June 11, 1980 identified an external corrosion problem with the Reactor Coolant Pump (RCP) studs. The 16 studs in the Byron Jackson pump design used at the Calvert Cliffs units hold the motor drive mount co the pump. The area where the studs exhibited corrosion is the surface exposed closest to the pump case extending up as high as 3-1/4" and into the pump case thread area on some studs.
All 64 RCP studs used in the 4 RCPs at CCNPP-2 were inspected during the Cycle 4 outage and 12 of these studs on RCPs 22A and 22B were affected (28).
BG&E has determined that the stud corrosion problem was caused by pump venting techniques, RCP seal pressure line failures and the presence of other metal objects which encouraged localized deterioration. They believe that improved pump venting techniques increased inspection frequencies, improved housekeeping techniques, and seal line modifications will eliminate the stud deterioration problem. We agrea with this conclusion. The I&E resident inspectors will confirm satisfactory replacement of all 12 RCP studs identified in Reference 28 prior to reactor startup from the Cycle 4 refueling outage.
2.10 Hydraulic Snubber Common Reservoirs
~During a routir.e surveillance inspection of all inaccessible snubbers at Calvert Cliffs-2 on October 8,1977, all eight (8) hydraulic snubbers supporting Steam Generator 22 were found to have insufficient hydraulic fluid. This was due to a single cracked fitting on snubber 2-63-19 and l
the installation which utilized a common reservoir for the 8 snubbers.
In the process of authorizing continued operation without monthly sur-veillance required on all inaccessible safety-related hydraulic snubbers, we requested the f011owing information for both Calvert Cliffs units:
1.
Provide a listing of all snubbers using a common reservoir. We believe this information should be available from installation drawings and need not wait for the next scheduled surveillance period as proposed by BG8E.
2.
Propose a program to either eliminate common reservoirs on safety related equipment or increase the surveillance of reservoir levels as a means I
of increasing assurance that snubbers will be highly reliable.
L
5 3.
Review the procentres used to implement the Operational Quality Assurance Program as it relates to minor construction or maintenance activities, such as that identified as the likely cause of this occurrence, in areas containinJ safety related equipment.
Identify and correbt any deficiences in these procedures.
This requested data was provided as follows:
No. Snubbers Reservoir Accessible Inaccessible Unj Unit 2 2
X 7
27 3
X 1
0 8
X 1
0 2
X 21 16 3
X 0
8 (SG)
X 2
2 9
X 1
0 BG&E has indicated their program to either eliminate common reservoirs in safety-relcted equipment or increase the surveillance of
<3ervoir levels as a means of increased assurance that snubbers will
.1 highly reliable.
BG&E has informed us that during the Cycle 4 refueling outage for Unit 2, they will modify all safety related snubbers to have their own separate hydraulic fluid reservoir, with the exception of the eight snubber groups supporting each steam generator (SG - indicated in the above tabla). We find these modifications, to all but the SG snubbers, acceptable.
SGEE has stated that Bechtel (the plant architectural engineeri, Grinnell (a snubber manufacturer) and their own staff do not recommend putting the eight SG snubbers on individual reservoirs.
Since BG&E has agreed to continue the monthly inspection of each reservoir level supplying the "ets of SG snubbers and has made such an application for change to TS Table 3.7-4, Reference 29, we conclude that a common mode failure of all snubbers supporting any SG'is unlikely. We find the proposed TS acceptable.
2.11' Reactor Coolant Pump Lube Oil Collection System In our October-2, 1980 letter on the status of the fire protection items at the Calvert Cliffs units, we concluded the RCP oil spillage protection system (Item 3.3.4) proposed by BG&E was acceptable. BG&E re~ ports that such an oil. collection system has been installed on each RCP during the Cycle 4 reload outage. The I&E resident inspector will confirm the accept-ability of this installation, thus closing out fire protection Item 3.3.4.
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> 2.12 Shutdown Cooling Relief Valve Testing In the application of January 29, 1981, BG&E proposed Change No. 4 to allow stopping both shutdown cooling (SDC) pumps for about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to allow the removal, testing and reinstallation of SDC relief valve RV-469 (30). Their proposal.was to change TS 3.10.5 to extend this special test exception to
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cover Mode 6 (refueling mode).
In discussions with BG&E, it was agreed that such a change could easily be added to TS 3.9.8 which is for Mode 6 coolant circulation and already has a special provision for deengergizing the SOC pumps for testing containment penetration No. 41.
However, we believe additional requirements on the water level above the top of the core and temperature monitoring are necessary.
BG&E has proposed that the testing of RV-469 only be performed when the water level is greater than 23 feet above the top of the irradiated fuel.
This is consistant with our request for TS change dated June 11, 1980 and is, therefore, acceptable.
3.0 Technical Specification Changes
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A brief description of each type of TS changes fo: the core refueling follows.
One of your proposed changes is not acceptable at this time. The intent of change Number 8 was to delete a numerical limit (21 kW/ft) below which fuel centerline melting will not occur.
As th6 basis for this revision, BG&E stated the deletion wss ta standardize the TS to other CE units.
We have reviewed TS for other facilitfes (e.g., Millstone, Unit No. 2; and the CE and the B&W Standard Technical Specifications) and found expl
- cit numerice.i limits for the centerline fuel melt. Though such a limit may not by in the. TS for all plants, we believe that the retention of the present 21.0 kW/i'. limit in the CCNP-2 TS bases is justified. This value is not readily found elsewhere and there is no other apparent operational limit to be used for demonstrating compliance with the specific acceptable fuel design limit for avoiding fuel centerline melting.
In discossion with the BG&E staff, it was decided to consider this proposed change withdrawn at this time.
If BG&E still desires this change, justification should ce provided with a new application for such a proposed TS change.
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, 3.1 Steam Generator Pressure Setpoints As an outcome of the review of a postulated steam line break, the following changes to the Technical Specifications were requested:
e Increase the steam generator low-pressura reactor scram setpoint from 478 psia to 570 psia, and e Increase the steam generator low-pressure trip bypass from below 600 psia to below 685 psia. This is a manual operator action, intended to bypass various control logics during normc1 ccoldown.
The pages affected are 2-9, 2-10, B2-5, 3/4 3-4, 3/4 3-15, and 3/4 3-17.
The reason for changing the steam generator reactor trip and bypass setpoints is to envelope the consequences of a steam line break for two cycles, such that reanalyses for the Cycle 5 operation will be bounded by the Cycle 4 analyses.
Based on bounding evaluations conducted by BG&E, the staff concludes that increasing the Technical Specification limits for the steam generator low-oressure setpoint to 570 psia and the steam generator bypass trip to below 685 psia will not adversely affect the consequences of SLB0C and therefore are acceptable.
3.2 Asymmetric Steam Generator Transient Protection Trip Function This new trip function would be covered in Table 2.2-1 requiring a steam gensrator pressure difference of < 135 psid. This trip function and the supporting analysis has been addressed in Sections 2.3 and 2.7 of this SE.
Therefore, we find the proposed TS acceptable. The pages affected are 2-9, B2-7, 3/4 3-2, 3/4 3-6 and 3/4 3-7.
3.3 Thermal Margin Limit Figure 2.1-1 should be modified to reflect the change in DNBR from 1.19 to.l.195.
In addition, pages B 2-1, B 2-3, B 2-5, B 2-6 and 3/4 2-2 should be 05anged to agree with the new DNBR limit found acceptable in Section 2.3 of this SE.
3.4 Axial Shape Index The axial shape index (ASI) should be changed slightlyT to accommodate the expected Cycle 4 flux peaking. The ASI was an input to the thermal margin safety limits analysis. Since this a.salysis has been reviewed and found acceptable, the proposed TS changes should be made. The pages affected are 2-11, -3/4 2-4 and 3/4 2-11.
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. 3.5 Shutdown Margin for Modes 1, 2, 3 and 4 Because of the higher fuel enrichment, the shutdown margin will need to be increased from 3.4% Ak/k for Cycle 3 /to 4.3% ok/k for Cycle 4 operation.
As mentioned iri Se'. tion 2.2 of this SE, we find the core reactivity analysis acceptable. The pages affected are 3/41-1 and B 3/41-1.
3.6 Shutdown Mar ll,i,n for Mode 5 The shutdown margin was evaluated for a boron dilution event during the cold shutdown condition.
It was de. ermined that a 3%21k/k snutdown margin would be required so that at least 15 minutes would be available to the operator in order to terminate the deboration transient. Level requirements are also proposed for the pressurizer and operating requirements for the RCPs. We find these proposed TS changes acceptable. The pages affected are 3/4 1-3, 3/4 1-9, 3/4 1-11, 3/4 1-13, 3/4 1-16, 3/4 4-2, 3/4 9-1, B 3/4 1-1 and B 3/4 1-2.
3.7 Concentration and Availability of Borated Water Sources Due to tFe required increase in the shutdown margin, the licensee proposes to increase the boron concentration from > 1720 ppm to > 2300 ppm and slightly reduce the storage requirements. We find that the change in the boron con-centration is sufficient to offset the increased core reactivity in acciaent analysis ard for shutdown margin control. The change is, therefore, acceptable. The pages.affected are 3/4 1-1, 3/4 1-3, 3/4 1-14, 3/4 1-15, 3/4 1 'l6, 3/4 5-1, 3/4 5-7,. 3/4 9-1, 3/4 10-1, B 3/4 1-2, B 3/4 1-3,' and B 3/4 9-1.
3.8 '4cderator Temperature Coefficient Tne change of -2.3 E 4 ok/k/ F to -2.2 E 4 ok/k/*F on page 3/4 1-5 is acceptable according to our SE, Section 2.2.2.
3.9 Augmentation Factor TS' Figure 4.2-1 augmentation factors have been increased to envelope future cycles. We find this not curve on page 3/4 2-5 accept 6ble.
3.10 ' Radial Peaking Factor factor (F ) have been changed fNm <nd total integrated radial peaking Planar radial. peaking factor (F 1a 1.61 to < 1.F3 and from < 1.54 to r
< 1.62, respectively. This change nas'been evaluated in SectTon 2.2 of the 3E and has been found acceptable. The TS pages changes are 3/4 2-6, 3/4 2-8 and 3/4 2-9.
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, 3.11 Containment Purge Isolation In Table 3.3-5 on page 3/4 3-20, the Containment Purge Isolation Valve Response Time has been reduced from 6 seconds to 5 seconds.
This change is in accordance with our present guidance (CSB Technical Position 6-4) and is, therefore, acceptable.
BG&E can meet this closure time reduction sin.ce the containment purge isolation valves have been modified to only partially open.
3.12 TSP Volume and Sample Method The proposed change would increase the minimum volume of trisodium phosphate dodecahydrate (TSP) required from 75 cubic feet to 100 cubic feet.
It would also change sample volume to 4.0 + 0.1 gms in >.5 + 1 liters of RWT water instead of 0.6 +.7 lbs in 80U gallons of water.
The reason for this change is that tWe minimum volume of TSP needed to raise the pH of the borated water of the ECCS to 7.0 is 100 cubic feet because of the boron concentration change (Section 3.7).
In order to test the ability of the TSP to raise the pH of the borated water of the ECCS, the ratio of the volume of TSP tc the volume of ECCS borated water must be the same in conta?nment as it is in the laboratory We find this change which is shown on TS page 3/4 5-5, acceptable, a.0 Physics Startup Testing he physics startup test program for Calvert Cliffs Unit No. 2, Cycle a was reviewed. The low power tests include CEA group worth, critical boron
- cncentration, isothermal temperature coefficient, and CEA symmetry check test. The power ascension test included power distribution, critical to-on concentration, isothermal temperature cctfficient, and power coeffi-
- ient tests. The acceptance criteria and review criteria for aach test are reasonable. The action and review plan section states the actions to be taken if any test fails to meet the acceptance or review criteria.
We have reviewed this entira program and find it acceptable.
1 5.0 Environmental Consideration ne have determined that this amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant enviror: mental impact.
Having made this determination, we have further concluded tnat the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)i ) that an environmental impact statement, 4
or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
. 6.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the prob-ability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable i
assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be con-ducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
February 10, 1981 s
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. ~ References 1.
Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company, to R. A. Clark, USNRC,
Subject:
Amendment to Operating License DPR-69, Fourth Cycle License Application, dated December 4,1980.
2.. Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company, to R. A. Clark, USNRC,
Subject:
Responses to NRC Staff Questions, dated January 30, 1981.
3.
Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company,
.to R. A. Clark, USNRC,
Subject:
Amendment to Operating License DPR-53, Fifth Cycle License Application, dated September 22, 1980.
4.
Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company, to R. A. Clark', USNRC,
Subject:
Amendment to Operating License DPR-53, Supplement 1 to Fifth Cycle License Application, dated Novemer 4,1980.
5.
Letter from R. A.- Clark, USNRC, to A. E. Lundvall, Jr., Baltimore Gas and Electric Company, dated December 12, 1980.
6.
" Fuel Evaluation Model," Combustion Engineering report CENPD-139-A, July 1974.
7.
Letter from O. D. Parr, USNRC, to F. M. Stern, Combustion Engineering, dated December 4, 1974.
8.- Letter from D. F. Ross. USNRC, to A. E. Scherer, Combustion Engineering, dated November 23, 1976.
9.
Letter from D. L. Ziemann, USNRC, to A. E. Lundvall, 'Jr., Baltimore Gas and Electric Company, dated November 23,.1976.
- 10. Letter from A. - E. Lundvall, Jr., Baltimore Gas and Electric Company, to D. L. Ziemann, USNRC, datad December 31, 1976.
- 11. Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company,
, to D. L. Ziemann, USNRC, dated February 23, 1977.
- 12. Memorandum from W.-Gammill and R. Vollmer, USNRC, to Operating Reactors Branch Chiefs, dated February 7,:1980.
13. Letter from A. E. Lundvall, Jr., Baltimore Gas arid Electric Compeny,
.to R. A. Clark, USNRC,
Subject:
Responses to NRC Staff Questions, dated November 20, 1980.
14.
CEPAN Method of Analyzing Creek Collgse of Oval Cladding," Combustion Engineering report CENPD-187 A, ' March 1976.
> 15. Memorandum from D. F. Ross and D. G. Eisenhut, USNRC, to D. B. Yassallo and K. R. Goller, " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Thermal Margin Calculations for Light Water Reactors," February 16, 1977.
- 16. PNO-77-221, preliminary notification of event on unusual occurrence of guide tube wear, December 14, 1979.
17.
Letter from A. E. Scherer, Combustion Engineering, to V. Stello, USNRC, dated December 23, 1977.
18.
Letter from W. Johnson, Mair.- Yankee Atomic Power Company, to V. Stello, USNRC, dated February 14, 1978.
- 19. Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company, to V. Stello, USNRC, dat. J February 17, 1978.
20.
Letter from R. A. Clark, USNRC, to W. G. Council, Northeast Nuclear Energy Company, dated October 6,1980.
- 21. Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company, to R. A. Clark, USNRC,
Subject:
CEA Guide Tube Inspection Program 5, dated January 21, 1981.
22.
Letter from A. E. Lundvall, Jr., Baltimore Ga and Electric Company, to R. A. Clark, USNRC,
Subject:
Sleeved CEA Guide Tube Evaluation Program,
. dated January 21, 1981.
23.
"Zircaloy Growth In-Reactor Dimensional Chunges in Zircaloy-4 Fuel Assemblies,"
Combustion Engineering report CENPD-198, December 1975.
24 "Zircaloy Growth Application of Zircaloy Irradiation Growth Correlations for the Calculation of Fuel Assembly and Fuel Rod Growth Allowances,"
Combustion Engineering. report CENPD-198, Supplement 1, December 1977.
25.
" Response to Request for Additional Information on CENPD-198-P, Supplement 1,"
Combustion Engineering report CENPD-198, Supplement 2-P, November 1,1978.
- 26. LLetter from R. L. Baer, USNRC, to A.- E. Scherer, Combustion Engineering, dated August 21, 1979.
- 27. Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company, to R. A. Clark, USNRC,
Subject:
Reactor Coolant System Vents, dated December 18, 1980.
28.
Letter from A. E.' Lundvall, Jr., Baltimore Gas and Electric Company, t1 R. A. - Clark, USNRC,
Subject:
Reactor Coolant Pump Stud Corrosion, dated February 5, 1981.
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D h 29. Letter from A. E. Lundvail, Jr., Baltin. ora Gas and Electric Company, to R. A. Clark, USNRC,
Subject:
Hydraulic Snubber Common Reservoirs, dated December 12, 1980.
30.
Letter from A. E. Lundvall, Jr., Baltimore Gas and Electric Company, to
- H. R. Denton, USNRC,
Subject:
Four Miscellaneous TS changes, dated
' January 29,1981.
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