ML20085M484
| ML20085M484 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/07/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20085M476 | List: |
| References | |
| NUDOCS 9506290213 | |
| Download: ML20085M484 (12) | |
Text
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t UNITED STATES j
NUCLEAR REGULATORY COMMISSION
'f WASHINGTON, D.C. 20555 0001
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EAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 104 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO.
103 T0 FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By letter of February 6,1995, as supplemented by letters dated March 23, and May 22, 1995, Pacific Gas and Electric Company (PG&E or the licensee) submitted a request for changes to the Technical Specifications (TS).
The proposed amendments would allow for the storage of fuel with an enrichment not to exceed a nominal 5.0 weight percent (wt%) U-235 in the new (fresh) and spent fuel storage racks.
The proposed changes would also clarify allowed substitution of fuel rods with filler rods and use of ZIRLO fuel cladding.
The licensee's supplemental letters provided additional clarifying information and did not change the initial no significant hazards consideration l
determination that was published in the Federal Reaister on March 1, 1995 (60 FR 11138).
2.0 EVALUATION The staff's evaluation of the criticality aspects of the proposed changes follows.
The licensee's submittals did not include a request to increase burnup of the fuel.
Fuel Enrichmgni The analysis of the reactivity effects of fuel storage in the new and spent fuel storage racks was performed with the three-dimensional multi-group Monte Carlo computer code, KENO-Sa, using neutron cross sections generated by the NITAWL code package from the 27 energy group SCALE data library.
Since the KENO-Sa code package does not have depletion capability, burnup analyses were performed with the two-dimensional transport theory code, CASM0-3.
CASM0-3 was also used to determine the reactivity effects of material and manufac-turing tolerances. These codes are widely used for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments.
These experiments simulate the Diablo Canyon fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment, assembly spacing, and absorber worth.
The intercomparison between two independent methods of analysis (KENO-Sa and CASMO-3) also provides an acceptable technique for validating calculational metheds for nuclear criticality safety. To minimize the statistical l
9506290213 950607 PDR ADOCK 05000275 P
y, uncertainty of the KENO-5a reactivity calculations, a minimum of 500,000 neutron histories were typically accumulated in each calculation.
Experience has shown that this number of histories is quite sufficient to assure convergence of KENO-Sa reactivity calculations. Based on the above, the staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the Diablo Canyon new and spent fuel storage racks with a high degree of confidence.
The fresh fuel storage vault contains two 5 x 7 arrays of storage locations with each array providing 35 cells arranged on a 22-inch lattice spacing. The two arrays are separated from each other by about 27.5 inches. The storage vault is intended for the receipt and storage of fresh fuel under dry (air) conditions.
However, to assure the criticality safety under normal and accident conditions and to conform to the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling, two separate criteria must be satisfied as defined in NRC Standard Review Plan (SRP), Section 9.1.1.
These criteria state that the maximum reactivity of the fully loaded fuel racks shall not exceed a k,,, of 0.95 if fully flooded with unborated water or a k of 0.98 assuming the optimum hypothetical low density moderation (e.g,,.,,fog or foam). The maximum calculated reactivity must include a margin for uncertainties in reactivity calculations and in manufacturingtolerancessuchthatthetruek'Sility,95percentconfidence will not exceed the calculated maximum value at a 95 percent prob level (95/95).
Since Diablo Canyon may contain Westinghouse standard or optimized (0FA) fuel designs with a 17 x 17 fuel rod array, calculations were performed to i
determine the more limiting fuel type from a reactivity standpoint. The Westinghouse OFA fuel is limiting in the fully flooded condition while the standard fuel exhibits the higher reactivity under low-density optimum moderation conditions. The maximum k for a fully loaded vault of 0FA fuel enriched to 5.0 wt% U-235 was calculaf'e'd to be 0.945 under fully flooded conditions.
For the hypothetical low-density optimum moderation condition, the maximum calculated k was 0.900 at a moderator density of approximately 8 percent of full densit'y',for a fully loaded vault of standard fuel. The calculations included a calculational bias and uncertainty derived from benchmark calculations, as well as uncertainties due to KENO-Sa statistics, lattice spacing, fuel enrichment, and fuel density at the 95/95 probability / confidence level. The results conform to the acceptance criteria of SRP Section 9.1.1 and are, therefore, acceptable.
The storage racks in the spent fuel pool are divided into two regions.
Region I contains 290 stainless steel storage cells with each cell surrounded on all four sides by Boraflex neutron absorber panels. The cells are spaced 10.93 inches apart with a 1.786 inch water flux-trap between two adjacent Boraflex panels.
Region 2 consists of 1034 storage cells and contains no Boraflex.
The cells are stainless steel with an inside dimension of 8.85 inches arranged on a 10.929-inch center-to-center spacing, providing a 1.899-inch water gap between the walls of the storage cells. The spent fuel racks are normally fully flooded by water borated to at least 2000 ppm of boron as required by the plant TS. However, to meet the criterion stated in
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SRP Section 9.1.2, k,,,ticipated reactivity and flooded with unborated water at must not exceed 0.95 with the racks fully loaded with fuel of the highest an
-a temperature corresponding to the highest reactivity. The maximum calculated reactivity must include a margin for uncertainties in reactivity calculations and in manufacturing tolerances such that the true k,,, will not exceed 0.95 at _a 95/95 probability / confidence level.
j Initial calculations for Region I have shown that 0FA fuel gave a slightly higher rack reactivity than the corresponding enrichment for standard Westinghouse fuel.. The spent fuel storage racks in Region I were reevaluated for 5.0 wt% U-235 enriched fuel moderated by pure water at 20*C with a density of 1.0 gs/cc, which results in the highest reactivity.
For the nominal storage cell design in Region 1, uncertainties due to tolerances in fuel-enrichment and density, fuel pellet diameter, storage cell inner diameter, j
stainless steel thickness, water gap thickness, Boraflex width and thickness, and boron-10-(B-10) loading-were accounted for as well as eccentric fuel positioning. These uncertainties were appropriately determined at the 95/95 probability / confidence level.
In addition, calculational and methodology biases and uncertainties due to benchmarking were included.
The reactivity calculations also considered the effects of Boraflex shrinkage and gap formation. All Boraflex panels were modeled with 4 percent' shrinkage.
i Because of the design of the racks, two different gap assumptions were made, depending on whether the Boraflex panel is located in the rack interior or the rack periphery. The interior panels are held in place by a stainless steel cover plate that is spot welded every 12 inches along each vertical edge through small cutouts in the Boraflex. Because of the localized stresses that would develop by these restraints due to a maximum shrinkage of 4 percent of
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the Boraflex panel in.the 12-inch interval, a gap of 0.48 inches was assumed j
to occur at the cutout location every 12 inches along the length of the panel.
In addition, all perimeter Boraflex panels were assumed to have a 14-inch gap located at the same axial location (top 14 inches).
Based on the results of blackness testing performed at other storage' facilities, and on upper bound I
values recommended by Electric Power Research Institute (EPRI), the staff concurs that these assumptions bound the current measured data and future development of additional shrinkage and gaps. The final Region 1 design, when fully loaded with fuel enriched to 4.5 wt% U-235, resulted in a k,, of 0.9421 g
when combined with all known uncertainties. This meets the staff s criterion of k no greater than 0.95 including all uncertainties at the 95/95 probbility/ confidence level and is, therefore, acceptable.-
To enable the storage of fuel assemblies with nominal enrichments greater than 4.5 wt% U-235, the concept of reactivity equivalencing was used.-
In this i
technique, which has been previously approved by the staff, credit is taken for the reactivity decrease due to the integral fuel burnable absorber (IFBA) material coated on the outside of the U0, pellet.
Based on these i
calculations, the. reactivity of the fuel rack array, when filled with fuel-assemblies enriched to 5.0 wt% U-235'with each containing 16 IFBA rods, was j
found to be 0.9444, thus meeting the acceptance criterion of 0.95.
The calculation assumed IFBA rods in the most reactive configuration with 2.25 mg/ inch per rod of B-10.
Fuel assemblies containing a nominal 36 mg/ inch j
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, of B-10 are equivalent to assemblies containing 1G IFBA rods at 2.25 mg/ inch per rod. Fuel assemblies containing a nominal 72 mg/ inch B-10 are equivalent to assemblies containing 32 IFBA rods at 2.25 mg/ inch per rod. The calculations included an uncertainty on the B-10 loading in IFBA rods.
As an alternative method for determining the acceptability of fuel storage in Region 1, the concept of burnup credit reactivity equivalencing was used.
This is predicated upon the reactivity decrease associated with fuel depletion and has been previously accepted by the staff for spent fuel storage analysis.
For burnup credit, a series of reactivity calculations are performed to i
generate a set of initial enrichment versus fuel assembly discharge burnup less than 0.95 when stored in ordered pairs which all yield an equivalent k ' Attachment E, Figure 1 of the the spent fuel storage racks. ThisisshownYn licensee's submittal dated February 6, 1995, in which a fresh 4.5 wt% enriched fuel assembly yields the same rack reactivity as an initially enriched 5.0 wt%
assembly depleted to approximately 3.73 MWD /Kgu.
A third alternative for storage of fuel assemblies enriched to 5.0 wt% U-235 and containing no IFBA rods in Region I consists of arranging the fuel in an l
alternating (checkerboard) configuration. Three configurations were analyzed; a checkerboard pattern of fuel assemblies and water-filled cells, a checkerboard pattern of fuel assemblies and assemblies with 32 IFBA rods l
(72 mg/ inch B-10), and a pattern with alternate rows of fuel assemblies and water-filled cells. The licensee has stated that calculations show that the reactivity of an assembly containing a nominal minimum of 72 mg/ inch B-10 is equivalent to the reactivity of a fuel assembly with 8,000 MWD /MTU cumulative burnup. The resulting 95/95 k values were 0.852, 0.944, and 0.895, respectively, all meeting the NC acceptance criterion of no greater than 0.95.
The Region 2 spent fuel storage racks were reanalyzed for storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 5.0 wt%
U-235 using the concept of burnup reactivity equivalencing.
For Region 2, the 1
Westinghouse standard fuel assembly design gave a slightly higher reactivity than the 0FA. The same initial assumptions, biases and uncertainties as used for the Region 1 analyses were included, except for the design basis temperature and the effects of Boraflex shrinkage and gaps.
Since the Region 2 racks contain no Boraflex, the temperature coefficient of reactivity is positive and a temperature of 150'F was assumed. A depletion uncertainty of 0.0005 times the burnup in MWD /KgU was assumed, resulting in an uncertainty of 0.02 Ak for fuel burned to 40 MWD /KgU.
This uncertainty is consistent with current practice and is acceptable. The equivalencing showed that fresh standard Westinghouse fuel enriched to 1.74 wt% U-235 yields the same rack reactivity (k 'sh fuel enriched to 1.79 wt% U-235 was equivalent to 5.0 wt% assemb For 0FA fuel irradiated to 38.75 MD/KgU, yielding a rack reactivity (k
) of 0.9462.
These values meet the NRC acceptance criterion of 0.95 and are,a,c,ceptable.
4 Fuel initially enriched to 5.0 wt% U-235 may also be stored in a checkerboard pattern in Region 2, alternating with cells filled with only water or non-fissile material.
For this case, the maximum calculated reactivity, including uncertainties, was 0.9392.
Most abnormal storage conditions will not result in an increase in the k,,, of the racks. However, it is possible to postulate events, such as the misloading of an assembly with an enrichment and burnup (or IFBA) combination outside of the acceptable area or pool temperatures exceeding 150 F, which could lead to an increase in reactivity for Region 2.
However, for such events credit may be taken for the presence of approximately 2000 ppm of boron i
in the pool water required by TS 3.9.14.2 since the staff does not require the i
assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (Double Contingency Principle). The reduction in k ',by credible accidents. caused by the boron more than offsets the reacti additioncauseEf In fact, the licensee has determined that only 400 ppm of boron is necessary to mitigate the worst postulated j
accident in any pool region. Therefore, the staff criterion of k,,, no greater than 0.95 for any postulated accident is met.
Use of Filler Rods In the event that a limited number of fuel rods in an assembly are damaged and cannot be replaced by similar fuel rods, the licensee has proposed using zirconium alloy or stainless steel filler rods with the requirement that the analyses for substituting the filler rods in fuel assemblies must be performed with codes and methods that have been approved by the NRC and must be demonstrated to comply with all fuel safety design bases. This is consistent with NRC Generic Letter 90-02, Supplement 1, " Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications,"
and is, therefore, acceptable.
Use of ZIRLO Claddino Another proposed change would allow the use of ZIRLO, in addition to Zircaloy-4, as an acceptable cladding material. ZIRLO is an improved zirconium-based fuel rod cladding material that has a lower corrosion rate and reduced radiation-induced growth. The staff has previously found ZIRLO to be acceptable and has revised 10 CFR 50.44 and 50.46 to include ZIRLO as an acceptable cladding material. Any use of ZIRLO clad fuel in the core will be evaluated using NRC-approved codes as part of the licensee's cycle-specific core reload safety evaluation. Therefore, this change is acceptable.
Technical Soecification Chanaes The following Technical Specification changes have been proposed as a result of the requested enrichment increase, as well as the proposed allowance for replacing fuel rods with filler rods, and the addition of ZIRLO as an acceptable fuel cladding. The staff finds these changes and the associated Bases changes acceptable.
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(1)
TS 3.9.14.1 and Figure 3.9-2 have been revised to allow the storage of spent fuel assemblies with initial enrichments up to 5.0 wt% U-235 in Region 2 of the spent fuel pool.
Fuel pellet diameters are considered in combination with initial enrichment and cumulative burnup to encompass both Westinghouse standard and 0FA fuel.
(2)
TS 3.9.14.3 and Figure 3.9-3 have been added to include the requirements for acceptable fuel storage in Region 1.
In addition, an action statement is included requiring suspension of all fuel movement and crane operations except to move the non-complying assemblies into an acceptable pattern.
(3)
TS 5.3.1 has been changed to remove reference to the number of fuel rods in each assembly, nominal length of each fuel rod, and maximum fuel enrichment.
In addition, the current allowance for fuel rod substitution as justified by analysis is being clarified to specify that the analysis be performed using NRC staff-approved methods, an allowance to use a limited number of lead test assemblies is being added, and ZIRLO fuel cladding is being allowed.
(4)
TS 5.6 has been changed to correct the word " borated" with "unborated" and to specify the maximum fuel enrichment allowed to be stored in the racks.
Based on the review described above, the staff finds the criticality aspects of the proposed enrichment increase to the Diablo Canyon new and spent fuel pool storage racks are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.
The proposed filler rod substitution and use of ZIRLO fuel rod cladding is also acceptable.
Although the Diablo Canyon TS have been modified to specify the above-mentioned fuel as acceptable for storage in the spent fuel racks, evaluations of reload core designs (using any enrichment) will, of course, be performed on a cycle-by-cycle basis as part of the reload safety evaluation process.
Each reload design is evaluated to confirm that the cycle core design adheres to the limits that exist in the accident analyses and TS to ensure that reactor operation is acceptable.
3.0 PUBLIC COMMENTS Public comments on the staff's proposed no significant hazards consideration determination (60 FR 11138) were provided by Jill ZamEk on behalf of the San Luis Obispo Mothers for Peace (MFP) by letter dated March 30, 1995.
The comments and staff responses follow:
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' Comment 1 The postulated fuel handling accident offsite thyroid doses could increase by a factor of 1.2.
MFP finds these calculations arbitrary, suspect, and not at all reassuring. MFP is looking for an increase in the marain of safety at the plant - not an increase in the risk factor. MFP finds this added risk a significant hazard and unacceptable.
Essoonse The licensee's submittal only requests an increase in fuel enrichment for new and spent fuel storage. No request has been made at this time to increase fuel burnup and, thus, radioactivity in individual fuel rods and the spent fuel pool will not increase due to this amendment.
Therefore, offsite thyroid doses from the postulated fuel handling accident will not change with the increase in fuel enrichment.
The licensee's discussion of the 1.2 factor increase in offsite thyroid doses resulting from a fuel handling accident refers to the bounding analysis contained in NUREG/CR-5009, " Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors," for up to 5.0 weight percent U-235 and 60,000 megawatt-days per metric ton Uranium (MWD /MTU) burnup. The licensee has not stated that offsite thyroid doses will increase by a factor of 1.2 but only that the consequences of an enrichment of up to 5.0 weight percent (wt%)
U-235 and an unchanged burnup will not approach the 10 CFR Part 100 values and are clearly bounded by the 1.2 factor (which included the higher burnup). The staff agrees with the licensee's conclusion, and further concludes that the offsite thyroid doses will not increase at all based on an increase in fuel enrichment alone.
While the licensee has not requested a higher burnup with this license amendment request, the staff anticipates that the licensee will subsequently submit such a request.
In that event, the staff will review the radiological impacts of higher burnup on all design basis accidents, including the fuel handling accident.
Comment 2 The seismic issue is one that continues to jeopardize the safe operation of the plant at DCNPP.
The recent earthquake in Japan has undeniably demonstrated that there exists no " earthquake-proof" structure. With the Hosgri Fault within a few miles of the plant, and other nearby faults, DCNPP is clearly vulnerable. MFP argues that increasing the radioactivity in the Spent Fuel Pools at the site unnecessarily increases the risks of a serious accident in the event of a seismic event.
Response
The increase in fuel enrichment (e.g., from 4.5 to 5.0 wt%) alone will not increase fission product inventory in fuel rods. Therefore, increased fuel enrichment will not increase radioactivity in the spent fuel pool.
It follows
,t ' ? o then that the risk of a seismic event is unchanged based on an increase in fuel enrichment alone.
Despite this, the staff wishes to correct unsupported conclusions in the comment regarding the Kobe earthquake and its implications for the Diablo Canyon Nuclear Power Plant (DCPP). DCPP was designed and constructed in the 1970's.
Since its original design, DCPP has undergone two extensive and thorough seismic reanalyses, the Hosgri reanalysis of the late 1970's and the Long-Term Seismic Program (LTSP) of the mid-1980's. The LTSP was performed in response to a license condition to conduct a comprehensive geosciences investigation. As part of the LTSP, a major seismic reassessment of the plant was conducted by the licensee and reviewed and approved by the NRC staff.
The scope of the Hosgri and LTSP assessments covered all the safety-related plant structures, systems and components including the spent fuel pool and the building. The spent fuel pool, which is founded on rock and constructed with thick (about 5 feet) reinforced concrete shear walls, is one of the most seismically rugged parts of the plant.
Both the Hosgri and LTSP reanalyses assumed the occurrence of a large (magnitude greater than 7) earthquake on the Hosgri fault at a distance of about 4 kilometers from the plant. The seismic demand used in these analyses was based on near-field data recorded from a number of large earthquakes. The reanalyses demonstrated that the seismic capacity of DCPP is greater than the demand of a large nearby earthquake with significant margins.
Most of the loss of life in the recent earthquake in Kobe, Japan was due to the collapse of residential structures that were not seismically designed.
Engineered structures that had earthquake damage were generally older and designed to codes that underestimated the size of the earthquake and its proximity to the city. Well engineered structures designed to more recent codes generally performed well with no significant structural damage.
For example, the new Kobe City Hall sustained no structural damage.
Based on the design and analyses of the DCPP and our review of developments in seismology and earthquake engineering, the NRC continues to have reasonable assurance as to the seismic adequacy of DCPP.
Comment 3 PG&E makes " analyses" to " verify" that an increase in the fuel enrichment would not involve a significant increase in the probability of (sic) consequences of an accident previously evaluated. MFP finds PG&E's assumptions questionable.
In its discussion of non-borated water, optimum-density aqueous foam and soluble boron, PG&E's figures of "below 0.88" are dangerously close to criticality - criticality being 1.0.
MFP is alarmed by this proposed reduction in the margin of safety.
Eft 1PEuif Normally, fresh fuel is stored temporarily in a dry environment in the new fuel storage vault pending transfer to the reactor core. Under these conditions, the reactivity of the storage racks when filled with fuel of the
.. highest allowed enrichment is extremely subcritical (k-eff is usually less than 0.50 as compared to 1.0 for a critical system). However, moderator may be introduced into the vault under abnormal situations, such as flooding or t
the introduction of foam or water mist (for example, as a result of fire fighting operations).
Foam or mist affects the neutron moderation in the array and can result in a peak in reactivity at low moderator density (called
" optimum" moderation). Therefore, the NRC requires that the criticality safety enalyses must address two independent accident conditions in conformance to General Design Criterion 62 of 10 CFR Part 50, Appendix A, which requires the prevention of criticality in fuel storage and handling.
These two analysis conditions are:
(a)
With the new fuel vault filled with fuel of the maximum permissible reactivity and flooded with pure water, the maximum k-effective shall not exceed 0.95, including mechanical and calculational uncertainties.
(b)
With the new fuel vault filled with fuel of the maximum permissible reactivity and containing moderator at the (low) density corresponding to optimum moderation, the maximum k-effective shall be less than 0.98, including mechanical and calculational uncertainties.
The reactivity of the new fuel vault containing 4.5 wt% fuel for the low density accident (analysis condition (b) above) resulted in a k-effective of 0.880. The new analysis with 5.0 wt% fuel resulted in an increase of k-effective to 0.900.
Both of these values are well below the NRC requirement of.k-effective no greater than 0.98.
Therefore, the PG&E analysis, which shows that k-effective remains below 0.98 for this optimum moderation condition, meets the NRC requirement and is not a reduction in the margin of safety, (i.e., defined as the difference between 0.98 and 1.0).
Comment 4 MFP finds that the proposed changes would create new hazards that have not been previously evaluated. MFP asserts that the increased radioactivity of the proposed fuel would impact not only the Spent Fuel Pools at DCNPP, but
" low" level radioactive waste and storage, transportation of this waste, and all future handling. Again, MFP finds these increased hazards significant and unacceptable.
Response
Handling, storage, and transportation of low-level radioactive waste are not affected by the increase in fuel enrichment. Based on surveys of operating reactors, the NRC staff has determined that core thermal power is a more accurate indicator of radioactive waste production than fuel enrichment or burnup. Generation of radioactive waste is also dependent on the transport paths from the reactor coolant system to the radioactive waste processing systems.
The NRC staff evaluates radioactive waste processing systems using a computer model based on core thermal power and transport paths. Therefore, changes in fuel enrichment or burnup do not alter the basis for staff
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- acceptance of the means of handling, storage, and transportatic, of radioactive waste.
Comment 5 If PG&E's amendment request were to be granted, the current 18-month cycle for refueling at DCNPP would be extended to up to 24 months, tiFP is concerned by I
this 6 month extension, because it lengthens the period for inspections, surveillances and maintenance for certain safety-related systems and equipment. Because of PG&E's unique rate settlement agreement (1988), PG&E gets paid only when it produces power. This provides PG&E with the incentive to postpone or rush maintenance in order to increase profits.
PG&E's most recent outage was completed in an industry record time of 35 days. The Nuclear Regulatory Commission (NRC) sited [ sic] 7 violations during this period, and also voiced concern regarding PG&E's rushed work and its severe cuts in staff:
... recent declining trends observed in housekeeping, engineering coordination with the plant, and procedural compliance have raised our concern. Additionally, your efforts to streamline your organization and reduce outage duration may further stress your safety programs." MFP asserts that PG&E's efforts to increase its profits jeopardizes safety. MFP further asserts that the results of the proposed changes in the TS for DCNPP would serve to augment an existing problematic situation and further threaten the safe operation of the plant.
Response
The amendment request in question only requests approval for storage of new and spent fuei with enrichment of up to 5 wt%.
There have been no other requests to date from the licensee that support extended cycles.
However, the staff does anticipate that the licensee will submit such a request at a future date and for that reason we will address this comment.
The NRC issued Generic Letter (GL) 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," on April 2, 1991. This GL provides guidance to licensees for the preparation of amendment requests to support 24-month fuel cycles.
The staff has generically evaluated changes from 18-to 24-month surveillances and has found that the safety impact is small due to redundant components in safety systems and other means to demonstrate during operation that components remain operable. While the staff has found the impact to be small in general, each licensee must perform a technical evaluation which supports this conclusion for the given facility. Also, licensees must demonstrate on a case by case basis that plant component histories based on surveillance and maintenance data support the conclusion that the safety effect is small.
Licensees must also show that assumptions in the plant licensing basis remain valid based on an extended surveillance interval. The licensee's evaluation would include an assessment of increased calibration intervals and their effect on instrument errors to ensure that instrument drift will not result in errors that exceed assumptions of the safety analysis. The staff will review the licensee's supporting information to any proposed amendment request to 1
L g p 's increase the length of fuel cycles to ensure that the proposed changes do not have a significant effect on safety and will only approve amendment requests that are consistent with that conclusion.
In addition, the staff has identified certain benefits associated with extended surveillance intervals.
For instance, less frequent testing reduces component wear which, on balance, tends to increase system reliability. On this basis, the staff recommended certain changes to surveillance requirements in GL 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation." While this GL addressed surveillance testing during power operation, similar considerations apply to surveillance requirements conducted during refueling outages.
Likewise, a significant portion of maintenance can be performed while the unit is on-line.
The impact on safety of delaying for six months maintenance that can only be done during shutdown is small.
In addition, starting July 10, 1996, licensees must meet 10 CFR 50.65, " Requirements for monitoring the effectiveness of maintenance at nuclear power plants," which requires that important equipment is maintained in accordance with licensee goals such that it can be reasonably assured of performing as required. The maintenance rule is not inconsistent with 24-month cycles.
In our Systematic Assessment of Licensee Performance (SALP) report for DCPP dated September 30, 1994, within the context of our addressing the licensee's overall superior performance the staff mentioned that certain trends were of concern. Our cover letter alerted the licens,' to these areas and encouraged them to focus their attention in these areas to " assure continued superior safety performance." These are areas which the NRC will also continue to monitor to ensure that safety performance does not become unsatisfactory.
Conclusion The NRC has considered MFP's comments and has concluded that there is nothing in them that would cause the staff to change the proposed no significant hazards consideration determination.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Reaister on (60 FR 30120).
In this finding, the Commission determined that issuance of this amendment would not have a significant effect of the quality of the human environment.
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6.0 CONCLUSION
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The Commission has concluded, based on the considerations discussed above, j
that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, i
and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
L. Kopp M. Miller l
Date: June 7, 1995 l
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