ML20211K114

Proposed Tech Specs,Increasing Fuel Enrichment Limit for Cycle 2 & Subsequent Cycles & Revising Uncertainty Allowance for Spent Fuel Storage Racks

ML20211K114
Person / Time
Site:	Waterford
Issue date:	06/24/1986
From:	LOUISIANA POWER & LIGHT CO.
To:
Shared Package
ML20206J725	List: ML20211K114 W3P86-2163, Application for Amend to License NPF-38,changing Tech Specs to Increase Fuel Enrichment for Cycle 2 Limit & Revising Uncertainty Allowance for Spent Fuel Storage Racks.Affidavit Encl.Fee Paid
References
NUDOCS 8606270254
Download: ML20211K114 (80)

Text

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POTENTIAL TECHNICAL SPECIFICATIONS CHANGES Revision 0,.s/24/86' (3) GROUP A ( /01/86) (2) CROUP B ( ! / ( 9/25/86) (10/01/86)

(10/01/86) (11/21/86) (7) GROUP C (5) GROUP D TECH SPEC (12/25/86) .(Oy 01/87)

NATURE OF TECH SPEC NATURE OF TECH SPEC NATURE OF TECH SPEC NATURE OF SECTION CHANGE SECTION CHANGE SECTION CHANGE SECTION CHANGE i

_3.1.3.4 CEA Drop Time 3.1.2.7 May Require Change 3.1.1.1 Shutdown Margin for 2.1.1.1 Redefine the Limit, 18 month Surveillance in BAM Tank MODES 1-4 may change Reactor Core DNBR (Submitted to NRC Concentration on 6/24/86) 5.3.1 Fuel enrichment limit 3.9.1 Refueling Boron 3.1.1.2 Shutdown Margin for 2.2.1 DNBR trip limit increases from 3.7 Concentration May MODE 5 may change w/o to = >4.10 w/o Change (Submitted to NRC on 6/24/86) 5.6.1 Update to reflect new 3.1.2.2 See MODE 5 Shutdown 3.2.4 fuel storage criti-Revise Figure and Change Margin Change Format, DNBR Margin '

cality analyres for 4

Cycle 2 (Submitted to NRC on 6/24/86) l; 3.1.2.4 See MODE 5 Shutdown 3.3.1 Table 3.3-1 ACTION 6b Margin Change may be revised,

f. Reactor Protective l

4 Instrumentation 1

3.1.2.6 See MODE 5 Shutdown 3.1.3.6 Revise Insertion Margin Change. Limit Figure 3.1.2.8 See MODE 5 Shutdown Margin Change 3.1.2.9 Table 3.1-1 may change, Shutdown margin i,

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r POTENTIAL TECHNICAL SPECIFICATIONS CHANGES Revision 0, 6/24/86 i

(7) GROUP E ( 7/15/86) (7) 01/86) (12/22/85)

(12/15/86) GROUP F (1) GROUP G (8) CROUP H (12/15/86) (_6/22/86)

TECH SPEC NATURE OF TECH SPEC NATURE OF

} TECH SPEC NATURE OF TECH SPEC NATURE OF

. SECTION CHANGE SECTION CHANGE SECTION CHANGE SECTION CHANGE 3.1.1.3 'NTC will become more 2.1.1.2 Redefine the PLHR 2.2.2 Removal, requested, 3.3.l[3.3.2 RPS/ESFAS Surveillance negative at EOC and limit SIFR 10465, 3/26/86 (01/01/87) Interval more positive at BOC CPC Addressable constant i 3.1.3.7 Add curve and change 3.1.3.1 CEA Misalignment 3.6.3 Containment Isolation short-term and tran- ACTION statement (08/01/86) Valves sient insettion limits need to be modified

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3.2.5 Clarification, RCS 3.2.1 KW/FT Limit may 3.8.1.1 A.C. Sources -

flow rate change (11/01/86) DG Testing Relief 3.2.7 ASI ranges will 3.5.1 SIT boron concen- 6.2.2 Staffing change tration and volume (Submitted to NRC on requirements may 6/24/86) change Add RVLMS per License 3.3.3.6 3.5.4 RWSP boron concen- 3.4.8.2 PZR Spray Nozzle Usage l Conditions and CEOG tration and volume (08/01/86) Factor requirements may change ^

3 3.10.1 Surveillance Require- 3.10.1 ACTIONS a,b may 3.4.8 Pressure Vessel ment 4.10.1.2 may be have revised values (08/01/86) Surveillance Capsule relaxed, if desired of boron concentration 3.10.2 Add 3.1.3.7 for 3.3.3.8 Change Table 3.3.3.8 3.3.1 RPS, Allow Bypass l part-length CEAs to list smoke detector SG Level High Trip 1

in Control Room (Submitted to NRC on h 6/24/86)

5.3.1

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characterizing 1807 gr j

uranium as maximum fuel j rod loading (Submitted to NRC on i 6/24/86) i i

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-18 The following is a request to increase the level of enrichment for the fuel storage areas (spent fuel pool, new fuel storage vault, and ccatainment temporary storage racks) from 3.70 weight percent U-235 to 4.10 weight percent U-235.

Existing Specification See Attachment A Proposed Specification See Attachment B Description According to Technical Specification 5.3.1, the Waterford 3 fuel is limited to a maximum enrichment of 3.70 weight percent U-235. Because Cycle 2 is being designed as an approximately 18 month cycle, increased fuel enrichments are needed. For cycle 2 the maximum nominal enrichment will be 3.90 weight percent U-235, however, it is estimated that later cycles will require a maximum fuel enrichment of approximately 4.10 weight percent U-235.

Middle South Services, Inc. has performed a criticality analysis for each of the fuel storage areas (new fuel vault, spent fuel pool, and containment temporary storage racks) using KENO, a 3-D monte carlo criticality analysis code. Based upon these analyses, the resultant K-eff for each of these areas is less than the required limit of 0.95 for enrichments up to 4.10 weight percent U-235. A discussion of the methodology used and the uncertainties applied is included in Attachment C.

Approval of the proposed change is requested by October 1, 1986 in order to support fuel receipt for Cycle 2.

Safety Analysis The proposed changes described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed change will increase the fuel enrichment limit in order to allow receipt of reload fuel for use in extended cycle operation.

Because the calculated K-eff values (including uncertainties) indicate that the fuel storage configurations are substantially sub-critical, the probability of a criticality event in these areas is not increased. No physical change is being made to the storage areas.

Since a criticality event is demonstrated to be unfeasible, there are no increased adverse consequences for such a postulated event.

2. Will operation of the facility in accordance with the proposed change create the possibility of a new or dif ferent kind of accident from any accident previously evaluated?

Response: No. ,

a Because no physical change is being made to the facility, and because there will not be a change in how the facility is operated, the

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. proposed change will not introduce a new or different kind of accident from those previously evaluated.

3. Will operation of this facility in accordance with this proposed change involve a significant reduction in margin of safety?

Response: No.

Because the calculated values for K-eff (including uncertainties) are below the regulatory limits and because they reflect a substantial sub-critical configuration for each of the fuel storage areas under 1 adverse conditions, the margin of safety is not reduced by implementing the proposed change.

i' Safety and Significant Hazards Determination j Based upon the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined

, by 10 CFR 50.92; (2) there is a reasonable assurance that the health and j safety of the public will not be endangered by the proposed change; and (3) i this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement, i

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NPF-38-18 ATTACIDENT A t

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DESIGN FEATURES

.. ,, i 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 236 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of 1807 grams uranium. The initial core loading shall have a maximum enrichment of 2.91 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.70 weight percent U-235.

CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 83 full-length and 8 part-length control element assemblies.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements,

b. For a pressure of 2500 psia, and

c. For a temperature of 650*F, except for the pressurizer and surge line which is 700*F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 11,800

+600, -0 cubic feet at a nominal T,yg of 582.1*F.

5.5 METEOROLOGICAL TOWERS LOCATION

, 5.5.1 The primary and backup meteorological towers shall be located as shown on Figure 5.1-1.

WATERFORD - UNIT 3 5-5 l

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I NPF-38-18 ATTACHMENT B I

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1 pm kh DESIGN FEATURES 5.3 REACTOR CORE 1

FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel

assembly containing a maximum of 236 fuel rods clad with Zircaloy-4. Each

! fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of 1807 grams uranium. The initial core loading shall have a maximum enrichment of 2.91 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of +:9e weight percent U-235. l 4.10 CONTROL ELEMENT ASSEMBLIES

5.3.2 The reactor core shall contain 83 full-length and 8 part-length j control element assemblies.

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i 5.4 REACTOR COOLANT SYSTEM

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DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the

. applicable Surveillance Requirements, i

j b. For a pressure of 2500 psia, and

c. For a temperature of 650*F, except for the pressurizer and surge line

! which is 700*F.

VOLUME

) 5.4.2 The total water and steam volume of the reactor coolant system is 11,800

+600, -0 cubic feet at a nominal T,yg of 582.1'F.

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5.5 METEOROLOGICAL TOWERS LOCATION 5.5.1 The primary and backup meteorological towers shall be located as shown i

on Figure 5.1-1.

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J NPF-38-18 ATTACHMENT C

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WSES-3 FUEL STORAGE RACKS UPGRADE FOR THE STORAGE OF 4.1 WEIGHT PERCENT U-235 ASSEMBLIES bY M. R. EASTBURN j Middle South Services, Inc.

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l Abstract This report documents the criticality analyses of the WSES-3 spent fuel storage rack, the fresh fuel storage rack and the containment temporary storage rack. The FSAR states that it is safe to store fuel assemblies of up to 3.5 weight percent U-235 in all racks.

The purpose of this study is to determine if the multiplication factors of the racks are less than allowed limits when they are fully loaded with 4.1 weight percent U-235 fuel assemblies.

KENO, a Monte Carlo type code, was used to calculate the effective multiplication factors of conservative three-dimensional models of 1 the racks e.t the optimum water density of 1.0 gm/cc for the fresh fuel storage rack and at the prescribed density of 1.0 gm/cc for the other two racks.

A maximum K-EFF of 0.903 is obtained in the fresh fuel storage rack, which is less than the allowable limit of 0.95 for potential moderators and well below the limit of 0.98 for optimum modera-tion. The multiplication factor of the spent fuel and containment temporary storage racks at a water density of 1.0 gm/cc are 0.949 and 0.899 respectively, both below the allowable limit of 0.95.

From a criticality standpoint, 4.1 weight percent U-235 assemblies may be safely stored in all fuel storage racks. -

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1.0 Summary The FSAR shows that it is safe to store fuel assemblies of up to 3.5 weight percent U-235 in the fresh fuel storage rack, the spent fuel storage rack and the containment temporary storage rack.

This analysis shows that, from a criticality standpoint, assem-blies of up to 4.1 weight percent U-235 may be safely stored in the racks. It is shown that K-effective is lass than the NUREG-0800 limit of 0.95 for all racks.

2.0 Assumptions

1. The fuel racks are completely filled with unirradiated fuel

assemblies containing 4.1 weight percent U-235. The uranium dioxide density of the fuel stack is 10.061 grams / cubic cen-timeter (91.8 %TD)

2. The fresh fuel rack is moderated by pure water at various uniform densities. The spent fuel and containment temporary storage racks are moderated by pure water at a density of 1.0 grams / cubic centimeter.

3. No burnable poison, control element assembly or other fixed poison is stored with the fuel assembly.

4. The temperature of the rack, the concrete walls and all compo-nents of the fuel assembly is 20 degrees celsius.

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5. All concrete walls are modeled as being 100 centimeters thick.

All rack pits are modeled as being uncovered.

6. Rack structural material is not modeled for the fresh fuel and containment temporary storage racks. Rack support material is not modeled for the spent fuel rack.

7. The fuel assembly model comprises the fuel pins only. No as-sembly structural material is modeled.

3.0 Nuclear Methods 3.1 Computer Codes Used KENO-IV/S was used for all criticality calculations. A 123 group GAM-THERMOS cross-section library which had been processed by NITAWL-S to account for resonance self-shielding was used by KENO in all calculations. Descriptions of KENO-IV/S, NITAWL-S and both neutron libraries are contained in Reference 1.

3.2 Verification of KENO Methodology Five critical experiments performed by Babcock and Wilcox were analyzed with KENO using a 123 group GAM-THERMOS library which had been processed by NITAWL to obtain resonance corrected cross-sections. These experiments, details of which may be found in

Reference 2, utilized water moderated, aluminum clad, 2.46 weight percent U-235 fuel rods in various configurations. The five MSS KENO calculations give a mean K-effective of 0.9964 1 0.0082,

which is in good agreement with the twenty one criticals from Reference 2 which yield a mean K-effective of 0.9967 1 0.0087.

, Good agreement is also obtained when the MSS results are compared to the 123 group data from the seventy critical experiments in Reference 3 which yield a mean K-effective of 0.9958 1 0.0087.

The distribution of K-effectives of all three calculations was verified to be normal using SAS (Reference 4). The equivalency between the five MSS KENO benchmark cases and the ninety one com-bined ORNL and B&W KENO analyses was demonstrated using the F, T, and Bartlett tests, which have also been incorporated into SAS.

Therefore, using the mean and the standard distribution from the combined data, the KENO reliability factor is calculated as:

0.004 + (1.942)(0.008689) = 0.0209 I

where a one-sided tolerance factor of 1.942 has been applied (95%

probability at a 95% confidence level for ninety one samples, c.f.

Reference 5, Table 1.4.2). The KENO calculations will be correc-ted as follows for safety analysis:

I I K-eff(safety) = K-eff(KENO) + 0.021 l 5

No specific correction has been made for experimental uncertain-ties in the critical experiments cited in Reference 2 since they

are approximately ten times smaller than the statistical varia-tions in the calculational results. In accordance with Reference 6, no corrections due to differences between low density foam l

conditions and full density water conditions is needed.

3.3 Corrections to KENO results In addition to the 0.021 safety factor of paragraph 3.2, several other factors are added, where appropriate, to the KENO results to produce the final multiplication factor. These factors, taken from Table 9.1-8, Reference 7, are: 0.0012 for pellet den-l sification, 0.0001 for stainless steel dimensional tolerances and 0.0020 for Boraflex Boron-10 loading uncertainty. Other items in Table 9.1-8, Reference 7 have been explicitly accounted for in the i

models used in this analysis.

4.0 Results l

, 4.1 The Containment Temporary Storage Rack (CTSR)

The CTSR consists of a linear array of five cavities on eighteen inch centers whose internal dimensions are 8.62 inches on a side.

Since the outside assembly dimensions are 8.096 inches on a side, there is a 1 0.262 inch pitch-to-boundary uncertainty in the 1

lateral placement of any one assembly in its cavity. Both a 4

nominal and an adverse geometry were modeled. The nominal I

geometry had all assemblies centered in the storage cavities while l in the advers's geometry the assemblies were shifted within the cavities to minimize the distance to the center assembly. A NITAWL/ KENO analysis of the two geometries demonstrated that the j adverse geometry was the most reactive. Since this rack is i

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designed for wet storage of fuel, the analysis was performed with a water density of 1.0 gram per cubic centimeter.

The K-effective calculated from 22,000 neutron histories using the 123 group library is 0.87654 1 0.00519. After adding the KENO safety factor of 0.021 and a pellet densification factor of 0.0012, we obtain 0.899 as the multiplication factor of the rack l at the 95/95 confidence level.

4.2 The Fresh Fuel Storage Rack (FFSR)

There are eighty assemblies in the fresh fuel storage rack ar-ranged in eight groups of ten assemblies. Within each group the assemblies are arranged in two columns of five assemblies with a nominal spacing of twenty one inches center-to-center between the columns and between the assemblies in each column . There are four of these groups separated by forty nine inches center-to-center in each of two rows. The rows are separated by 58 inches center-to-center. Overall the assemblies form a eight by ten ar-ray. The rack design provides a 8.938 inch square cavity for each assembly. Since the assembly is 8.096 inches square pitch-to-pitch, there is an 1 0.421 inch uncertainty in the assembly placement.

A nominal and two adverse geometries were modeled. The nominal geometry had all assemblies centered in the storage cavities while l

in the adverso geometry the assemblies were shifted within the I storage cavities to either minimize the distance to the center of 1

each quadrant of the rack or to minimize the distance to the cen-ter of the entire rack. The adverse geometry therefore decreased the overall rack dimensions by 0.842 inches in each direction.

KENO calculations were made at several low densities until the low density peak reactivity was located. A KENO calculation was also made at 1.0 gm/cc. The maximum reactivity was obtained with full density water in the nominal geometry as shown by the KENO results in Table 1, below.

Table 1: KENO Results KENO RUN K-EFF

1. Nominal FFSR at 0.04 gm/cc 0.79706 1 0.00465

2. Nominal FFSR at 0.05 gm/cc 0.80739 i O.00493

3. Nominal FFSR at 0.06 gm/cc 0.79119 1 0.00479

4. Nominal FFSR at 1.00 gm/cc O.88095 1 0.00409

5. Adverse FFSR #1 at 0.05 gm/cc 0.78433 1 0.00486

6. Adverse FFSR #1 at 0.06 gm/cc 0.79853 1 0.00455

7. Adverse FFSR #1 at 0.07 gm/cc 0.79511 1 0.00491

8. Adverse FFSR #1 at 1.00 gm/cc 0.87535 i 0.00370 i .

l 9. Adverse FFSR #2 at 0.06 gm/cc 0.79178 1 0.00447 Based on 22,000 neutron histories the maximum multiplication fac-tor of the FFSR is 0.903 after the application of a 0.021 KENO safety factor and a 0.0012 pellet densification factor.

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4.3 Spent Fuel Storage Rack (SFSR)

The SFSR is an 32 by 34 array of 10.329 inch (minimum) square cans. Each can contains a 8.57 -0,+0.084 inch cavity for the fuel assembly and two flux traps on adjacent faces. The flux traps each contain two poison shims made of a 0.1 inch thick sheet of Boraflex sandwiched between two thin stainless steel sheets. A small water gap separates the Boraflex assemblies from the walls of the trap. A water gap 1.096" (min) to 1.116" (max) separates the two Boraflex sandwiches. The poison assembly is centered on the assembly and is one foot shorter than the active fuel and one inch narrower than the assembly. A six inch layer of rack was modeled without Boraflex on both the top and bottom and the Boraflex was modeled as 14 fuel pin pitches wide, 1.012 inches less than the fuel width. A 16X17 array of the basic cells, one quarter of the total 1088 assemblies, was modeled. In the final model the minimum trap size as well as the minimum can size was used to maximize the reactivity.

The restrictions of the complex KENO model made it impossible to model the adverse geometry using that model as a base. A sim-plified cell containing only fuel, Boraflex and water was used to calculate an adverse geometry factor. The nominal geometry of the

' simplified cell had the fuel centered while in the adverse geometry the fuel assembly was offset toward the lower right cor-ner as far as the 0.237 inch placement uncertainty would allow.

Both simplified cells were run in an radially infinite array and the difference in multiplication factors was used as an additive

factor to the result from the full model with the fuel centered in the cavity. Since this rack is intended for wet storage, a water density of 1.0 gm/cc was used for all cases.

1 The K-EFF calculated from 97,000 neutron histories using the nominal geometry is 0.91771 1 0.00235. After applying the KENO safety factors, totaling 0.021, the 0.0012 pellet densification factor, a 0.0001 SS304 machining tolerance factor, a 0.002 factor for Boraflex B10 loading uncertainty and an adverse geometry fac-tor of 0.0067 obtained from the simplified cell model, we obtain 0.949 as the K-eff of the SFSR at the 95/95 confidence level.

4.4 Results Summary After applying KENO safety factors and model uncertainties to the KENO results we obtain the values in Table 2 as the maximum mul-tiplication factors of the fuel storage racks at the 95/95 confidence level:

Table 2: Analysis Results RACK K-EFF

1. Adverse CTSR at 1.00 gm/cc 0.899

2. Adverse SFSR at 1.00 gm/cc 0.949

3. Nominal FFSR at 1.00 gm/cc 0.903 l

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4.5 Conservatisms

1. No credit was taken for neutron absorption in the upper or lower end fittings or grids in the assemblies, rebar in the concrete or structural support steel in the racks.

2. All concrete walls have been modeled as being 100h centimeters thick.

3. Fire fighting foam was modeled as pure low density water. No credit was taken for any added neutron absorbers.

4. Moderator was assumed to be 750 centimeters thick above the FFSR. This is sufficient to provide the equivalent of 30 cen-timeters of full density water in the 0.04 gm/cc low density foam case.

5. A system temperature of 20 degrees Celsius was used in the analysis of all racks. The densities of the fuel and moderator are greater and therefore more reactive than at j higher temperatures.

5.0 Conclusions This analysis *has shown that the WSES-3 fuel storage racks can, from a criticality standpoint, safely store 4.1 weight percent enriched fuel assemblies. Assurance is provided that X-effective

, is maintained below the calculational limit of 0.95 for the spent l

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fuel and containment temporary storage racks as required by NUREG-0800. The multiplication factor of the fresh fuel rack is less than the calculational limit of 0.95 for possible moderation and well below the 0.98 limit at optimum moderation, also as required by NUREG-0800. The design of all racks analyzed complies with the requirements of NUREG-0800 and either ANSI /ANS-57.2-1983 or ANSI /ANS-57.3-1983 as appropriate.

6.0 References

1. SCALE-2: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation. Oak Ridge National Laboratory, Doc. No. CCC-450, MSS File No. 015-27.

2. N. M. Baldwin, et. al. Critical Experiments Supporting Close i Proximity Water Storage of Power Reactor Fuel. BAW-1487-7, July 1979.

3. R. M. Westfall, J. R. Knight, " Scale System Cross-Section

Validation with Shipping-Cask Critical Experiments," Trans.

I Am. Nucl. Soc. 33, 368 (1979).

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4. ANSI N15.15-1974, American National Standard Assesment of the Assumption of Normality (employing Individual Observed values).

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5. R. E. Oden and D. B. Owen, Tables for Normal Tolerance 1

Limits, Sampling Plans and Screening. Marcel Dekker, Inc.,

New York, N.Y., 1980.

6. D. G. Napolitano, D. R. Harris, et. al., " Validation of the NITAWL-KENO Methodology in Modeling New-fuel Storage

{ Criticality," Trans. Am. Nucl. Soc. 44, 291 (1983).

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7. WSES-3 Final Safety Analysis Report, Amendment 7.

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-19 The following is a request to remove the delta K-eff allowance value and the reference to the FSAR in Technical Specification 5.6.1. This change is related to the increased fuel enrichments required for extended cycle operation (reference Proposed Change NPF-38-18).

Existing Specification See Attachment A Proposed Specification See Attachment B Description According to Technical Specification 5.6.1, K-eff is to be equivalent to less than 0.95 when flooded with unborated water, which includes a conservative allowance of 0.0455 delta K-eff for uncertainties as described in Section 9.1.2 and Table 9.1-8 of the FSAR. The conservative allowance value is specific to the criticality analyses currently documented in the FSAR. In conjunction with Proposed Change NPF-38-18, Middle South Services, Inc. has performed a criticality analysis for the spent fuel pool using i

KENO, a 3-D Monte Carlo analysis code. Based upon this analysis, the l resultant K-eff is less than the required limit of 0.95 for enrichments up l to 4.10 weight percent U-235. A discussion of the methodology used and the I uncertainties applied is included in Attachment C.

Since the allowance for uncertainties has been changed as a result of the l new analysis, it is necessary to modify the wording of this technical l specification. The reference to the FSAR is also being deleted to avoid l confusion as the Attachment C analysis will not be included in the

! Waterford FSAR until December, 1987 in accordance with 10CFR50.71(e).

Approval of the proposed change is requested by October 1, 1986 in order to l support fuel receipt for Cycle 2.

l l Safety Analysis The proposed changes described above shall be deemed to involve a i significant hazards consideration if there is a positive finding in any of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a significant increast in the probability or consequences of any accident previously evaluated?

Response: No.

Because the calculated K-eff values (including uncertainties) indicate that the fuel storage configurations are substantially sub-critical, the probability of a criticality event in these areas is not increased. No physical change is being made to the storage areas.

Since a criticality event is demonstrated to be unfeasible, there are no increased adverse consequences for such a postulated event.

2. Will operation of the facility in accordance with the proposed change create the possibility of a new or dif ferent kind of accident from any accident previously avaluated?

Response: No.

Because no physical change is being made to the facility, and because there will not be a change in how the facility is operated, the proposed change will not introduce a new or dif ferent kind of accident from those previously evaluated.

3. Will operation of this f acility in accordance with this proposed change involve a significant reduction in margin of safety?

Response: No.

Because the calculated values for K-eff (including uncertainties) are below the regulatory limits and because they reflect a substantial sub-critical configuration for each of the fuel storage areas under adverse conditions, the margin of safety is not reduced by implementing the proposed change.

Safety and Significant Hazards Determination Based upon the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined  ;

by 10 CFR 50.92; (2) there is a reasonable assurance that the health and l safety of the public will not be endangered by the proposed change; and (3) t this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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NPF-38-19 ATTACHMENT A i

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l DESIGN FEATURES

5. 6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with:

a. A k,9 f equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 0.0455 delta k for uncertainties as described in Section 9.1.2 and Table 9'$I8 of the FSAR.

b. A nominal 10.38 inch center-to-center distance between fuel assemblies placed in the spent fuel storage racks.

5.6.2 The k,ff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.3 The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation +40.0 MSL.

CAPACITY 5.6.4 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1088 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

WATERFORD - UNIT 3 5-6

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NPF-38-19 ATTACHMENT B i

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DESIGN FEATURES I

5.6 FUEL STORAGE CRITICALITY l

5.6.1 The spent fuel storage racks are designed and shall be maintained with:

a. Ak df equivalent to less than or equal to 0.95 when flooded with i

unborated water, which includes a conservative allowance ' ^ ^"'

de't: E for uncertainties,::'d::; rib d 17 S::ti:n ".1.2 :nd Toble dfC e' the TSA;h

b. A nominal 10.38 inch center-to-center distance between fuel assemblies placed in the spent fuel storage racks.

5.6.2 The k df f r new fuel f r the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.3 The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation +40.0 MSL.

1 CAPACITY 5.6.4 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1088 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

WATERFORD - UNIT 3 5-6

a - A A. ,a-- 4- - -- A - --a--- --- :-4 .-aa r

NPF-38-19 ATTACHMENT C

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File 304-37 WSES-3 FUEL STORAGE RACKS UPGRADE FOR THE STORAGE OF 4.1 WEIGHT PERCENT U-235 ASSEMBLIES by M. R. EASTBURN i

Middle South Services, Inc.

June 13, 1986 .

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2 Abstract This report documents the criticality analyses of the WSES-3 spent fuel storage rack, the fresh fuel storage rack and the containment temporary storage rack. The FSAR states that it is safe to store fuel assemblies of up to 3.5 weight percent U-235 in all racks.

The purpose of this study is to determine if the multiplication factors of the racks are less than allowed limits when they are i

fully loaded with 4.1 weight percent U-235 fuel assemblies.

KENO, a Monte Carlo type code, was used to calculate the effective multiplication factors of conservative three-dimensional models of the racks at the optimum water density of 1.0 gm/cc for the fresh fuel storage rack and at the prescribed density of 1.0 gm/cc for the other two racks.

A maximum K-EFF of 0.903 is obtained in the fresh fuel storage rack, which is less than the allowable limit of 0.95 for potential moderators and well below the limit of 0.98 for optimum modera-tion. The multiplication factor of the spent fuel and containment temporary storage racks at a water density of 1.0 gm/cc are 0.949 and 0.899 respectively, both below the allowable limit of 0.95.

From a criticality standpoint, 4.1 weight percent U-235 assemb;.ies may be safely stored in all fuel storage racks. -

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1.0 Summary I The FSAR shows that it is safe to store fuel assemblies of up to l 3.5 weight percent U-235 in the fresh fuel storage rack, the spent fuel storage rack and the containment temporary storage rack.

This analysis shows that, from a criticality standpoint, assem-blies of up to 4.1 weight percent U-235 may be safely stored in the racks. It is shown that K-effective is less than the NUREG-0800 limit of 0.95 for all racks.

2.0 Assumptions

1. The fuel racks are completely filled with unirradiated fuel assemblies containing 4.1 weight percent U-235. The uranium dioxide density of the fuel stack is 10.061 grams / cubic cen- I timeter (91.8 %TD)

2. The fresh fuel rack is moderated by pure water at various uniform densities. The spent fuel and containment temporary storage racks are moderated by pure water at a density of 1.0 grams / cubic centimeter.

3. No burnable poison, control element assembly or other fixed poison is stored with the fuel assembly.

4. The temperature of the rack, the concrete walls and all compo-nents of the fuel assembly is 20 degrees Celsius.

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5. All concrete walls are modeled as being 100 centimeters thick.

All rack pits are modeled as being uncovered.

6. Rack structural material is not modeled for the fresh fuel and containment temporary storage racks. Rack support material is not modeled for the spent fuel rack.

7. The fuel assembly model comprises the fuel pins only. No as-sembly structural material is modeled.

3.0 Nuclear Methods 3.1 Computer Codes Used KENO-IV/S was used for all criticality calculations. A 123 group GAM-THERMOS cross-section library which had been processed by NITAWL-S to account for resonance self-shielding was used by KENO in all calculations. Descriptions of KENO-IV/S, NITAWL-S and both neutron libraries are contained in Reference 1.

i 3.2 Verification of KENO Methodology

, Five critical experiments performed by Babcock and Wilcox were analyzed with KENO using a 123 group GAM, THERMOS library which had

'been processed by NITAWL to obtain resonance corrected cross-section's. These experiments, details of which may be found in Reference 2, utilized water moderated, aluminum clad, 2.46 weight percent U-235 fuel rods in various configurations. The five MSS KENO calculations give a mean K-effective of 0.9964 0.0082,

which is in good agreement with the twenty one criticals from Reference 2 which yield a mean K-effective of 0.9967 1 0.0087.

Good agreement is also obtained when the MSS results are compared to the 123 group data from the seventy critical experiments in Reference 3 which yield a mean K-effective of 0.9958 1 0.0087.

The distribution of K-effectives of all three calculations was verified to be normal using SAS (Reference 4). The equivalency between the five MSS KENO benchmark cases and the ninety one com-bined ORNL and B&W KENO analyses was demonstrated using the F, T, and Bartlett tests, which have also been incorporated into SAS.

Therefore, using the mean and the standard distribution from the combined data, the KENO reliability factor is calculated as:

0.004 + (1.942)(0.008689) = 0.0209 where a one-sided tolerance factor of 1.942 has been applied (95%

probability at a 95% confidence level for ninety one samples, c.f.

Reference 5, Table 1.4.2). The KENO calculations will be correc-ted as follows for safety analysis:

K-eff(safety) = K-eff(KENO) + 0.021 No specific correction has been made for experimental uncertain-ties in the critical experiments cited in Reference 2 since they are approximately ten times smaller than the statistical varia-tions in the calculational results. In accordance with Reference l 6, no corrections due to differences between low density foam i

__ ._ -. - - _ . = - -

O conditions and full density water conditions is needed.

3.3 Corrections to KENO results In addition to the 0.021 safety factor of paragraph 3.2, several other factors are added, where appropriate, to the KENO results to produce the final multiplication factor. These factors, taken from Table 9.1-8, Reference 7, are: 0.0012 for pellet den-sification, 0.0001 for stainless steel dimensional tolerances and 0.0020 for Boraflex Boron-10 loading uncertainty. Other items in Table 9.1-8, Reference 7 have been explicitly accounted for in the models used in this analysis.

4.0 Results 4.1 The Containment Temporary Storage Rack (CTSR)

The CTSR consists of a linear array of five cavities on eighteen inch centers whose internal dimensions are 8.62 inches on a side.

Since the outside assembly dimensions are 8.096 inches on a side, there is a + 0.262 inch pitch-to-boundary uncertainty in the lateral placement of any one assembly in its cavity. Both a nominal and an adverse geometry were modeled. The nominal geometry had all assemblies centered in the storage cavities while i

in the advers'a geometry the assemblies were shifted within.the

cavities to minimize the distance to the center assembly. A NITAWL/ KENO analysis of the two geometries demonstrated that the adverse geometry was the most reactive. Since this rack is l

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designed for wet storage of fuel, the analysis was performed with a water density of 1.0 gram per cubic centimeter.

l The K-effective calculated from 22,000 neutron histories using the 123 group library is 0.87654 1 0.00519. After adding the KENO safety factor of 0.021 and a pellet densification factor of 0.0012, we obtain 0.899 as the multiplication factor of the rack at the 95/95 confidence level.

1 4.2 The Fresh Fuel Storage Rack (FFSR)

There are eighty assemblies in the fresh fuel storage rack ar-ranged in eight groups of ten assemblies. Within each group the assemblies are arranged in two columns of five assemblies with a nominal spacing of twenty one inches center-to-center between the columns and between the assemblies in each column . There are four of these groups separated by forty nine inches center-to-center in each of two rows. The rows are separated by 58 inches center-to-center. Overall the assemblies form a eight by ten ar-ray. The rack design provides a 8.938 inch square cavity for each assembly. Since the assembly is 8.096 inches square pitch-to-pitch, there is an 0.421 inch uncertainty in the assembly placement.

A nominal and two adverse geometries were modeled. The nominal geometry had all assemblies centered in the storage cavities while in the adverse geometry the assemblies were shifted within the storage cavities to either minimize the distance to the center of each quadrant of the rack or to minimize the distance to the cen-ter of the entire rack. The adverse geometry therefore decreased the overall rack dimensions by 0.842 inches in each direction.

KENO calculations were made at several low densities until the low density peak reactivity was located. A KENO calculation was also made at 1.0 gm/cc. The maximum reactivity was obtained with full density water in the nominal geometry as shown by the KENO results in Table 1, below.

Table 1: KENO Results KENO RUN K-EFF

1. Nominal FFSR at 0.04 gm/cc 0.79706 1 0.00465

2. Nominal FFSR at 0.05 gm/cc 0.80739 1 0.00493

3. Nominal FFSR at 0.06 gm/cc 0.79119 1 0.00479

4. Nominal FFSR at 1.00 gm/cc 0.88095 1 0.00409

5. Adverse FFSR #1 at 0.05 gm/cc 0.78433 1 0.00486 1

6. Adverse FFSR #1 at 0.06 gm/cc 0.79853 1 0.00455 '

7. Adverse FFSR #1 at 0.07 gm/cc 0.79511 0.00491

8. Adverse FFSR #1 at 1.00 gm/cc 0.87535 1 0.00370

9. Adverse FFSR #2 at 0.06 gm/cc 0.79178 1 0.00447 Based on 22,000 neutron histories the maximum multiplication fac-tor of the FFSR is 0.903 after the application of a 0.021 KENO safety factor and a 0.0012 pellet densification factor.

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4.3 Spent Fuel Storage Rack (SFSR)

The SFSR is an 32 by 34 array of 10.329 inch (minimum) square cans. Each can contains a 8.57 -0,+0.084 inch cevity for the fua.1 assembly and two flux traps on adjacent faces. The flux traps each contain two poison shims made of a 0.1 inch thick sheet of Boraflex sandwiched between two thin stainless steel sheets. A small water gap separates the Boraflex assemblies from the walls of the trap. A water gap 1.096" (min) to 1.116" (max) separates the two Boraflex sandwiches. The poison assembly is centered on the assembly and is one foot shorter than the active fuel and one inch narrower than the assembly. A six inch layer of rack was modeled without Boraflex on both the top and bottom and the Boraflex was modeled as 14 fuel pin pitches wide, 1.012 inches less than the fuel width. A 16X17 array of the basic cells, one quarter of the total 1088 assemblies, was modeled. In the final model the minimum trap size as well as the minimum can size was used to maximize the reactivity.

The restrictions of the complex KENO model made it impossible,to model the adverse geometry using that model as a base. A sim-plified cell containing only fuel, Boraflex and water was used to calculate an adverse geometry factor. The nominal geometry of the

' simplified cell had the fuel centered while in the adverse geometry the fuel assembly was offset toward the lower right cor- i ner as far as the 0.237 inch placement uncertainty would allow.

Both simplified cells were run in an radially infinite array and the difference in multiplication factors was used as an additive 1

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factor to the result from the full model with the fuel centered in the cavity. Since this rack is intended for wet storage, a water density of 1.0 gm/cc was used for all cases.

The K-EFF calculated from 97,000 neutron histories using the nominal geometry is 0.91771 1 0.00235. After applying the KENO safety factors, totaling 0.021, the 0.0012 pellet densification factor, a 0.0001 SS304 machining tolerance factor, a 0.002 factor for Boraflex B10 loading uncertainty and an adverse geometry fac-tor of 0.0067 obtained from the simplified cell model, we obtain 0.949 as the K-eff of the SFSR at the 95/95 confidence level.

4.4 Results Summary After applying KENO safety factors and model uncertainties to the KENO results we obtain the values in Table 2 as the maximum mul-tiplication factors of the fuel storage racks at the 95/95 confidence level:

Table 2: Analysis Results RACK K-EFF

1. Adverse CTSR at 1.00 gm/cc 0.899

2. Adverse SFSR at 1.00 gm/cc 0.949

3. Nominal FFSR at 1.00 gm/cc O.903

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L 4.5 Conservatisms

1. No credit was taken for neutron absorption in the upper or lower end fittings or grids in the assembliss, rebar in the P

concrete or structural support steel in the racks.

2. All concrete walls have been modeled as being 100 centimeters thick.

3. Fire fighting foam was modeled as pure low density water. No

credit was taken for any added neutron absorbers. '

4. Moderator was assumed to be 750 centimeters thick above the J

FFSR. This is sufficient to provide the equivalent of 30 cen-timeters of full density water in the 0.04 gm/cc low density foam case.

5. A system temperature of 20 degrees Celsius was used in the analysis of all racks. The densities of the fuel and moderator are greater and therefore more reactive than at higher temperatures.

5.0 Conclusions This analysis 'has shown that the WSES-3 fuel storage racks can, from a criticality standpoint, safely store 4.1 weight percent i

a enriched fuel assemblies. Assurance is provided that K-effective is maintained below the calculational limit of 0.95 for the spent i

fuel and containment temporary storage racks as required by NUREG-0800. The multiplication factor of the fresh fuel rack is less than the calculational limit of 0.95 for possible moderation and well below the 0.98 limit at optimum moderation, also as required l by NUREG-0800. The design of all racks analyzed complies with the requirements of NUREG-0800 and either ANSI /ANS-57.2-1983 or ANSI /ANS-57.3-1983 as appropriate.

6.0 References

1. SCALE-2: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation. Oak Ridge National Laboratory, Doc. No. CCC-450, MSS File No. 015-27.

2. N. M. Baldwin, et. al. Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel. BAW-1487-7, July 1979.

3. R. M. Westfall, J. R. Knight, " Scale System Cross-Section Validation with Shipping-Cask Critical Experiments," Trans.

Am. Nucl. Soc. 33, 368 (1979).

4. ANSI N15.15-1974, American National Standard Assesment of the Assumption of Normality (employing Individual Observed Values).

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5. R. E. Oden and D. B. Owen, Tables for Normal Tolerance l

l Limits, Sampling Plans and Screening. Marcel Dekker, Inc., )

New York, N.Y., 1980.

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6. D. G. Napolitano, D. R. Harris, et. al., " Validation of the NITAWL-KENO Methodology in Modeling New-fuel Storage Criticality," Trans. Am. Nucl. Soc. 44, 291 (1983).

7. WSES-3 Final Safety Analysis Report, Amendment 7.

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NPF-38-20

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a DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-20 This is a request to revise Surveillance Requirement 4.1.3.4c, CEA Drop Time, by changing the surveillance interval from every 18 months to each refueling outage.

Existing Specification See Attachment A Proposed Specification See Attachment B Description The proposed change will revise Surveillance Requirement 4.1.3.4c by requiring CEA drop time measurements be performed at each refueling outage in lieu of every 18 months. This change will accommodate the present Waterford 3 Cycle 1 length and the extended cycle lengths for Cycle 2 and subsequent cycles.

Surveillance Requirement 4.1.3.4c was originally performed on 2/6/85. The 18 month interval is up on 8/6/86 with a late due date (as allowed by Surveillance Requirement 4.0.2a) of 12/20/86.

Waterford 3 is currently scheduled to begin the first refueling outage in mid November, 1986 and, therefore the late due date for Cycle 1 may be acceptable. However, should unforeseen circumstances occur the refueling outage could be extended beyond the late due date thus forcing a premature outage to perform Surveillance Requirement 4.1.3.4c. Additionally, Waterford 3 will be using a nominal 18 month Cycle 2 and may go to 24 month refueling cycles in the future. Even allowing the 25% interval extension from Surveillance Requirement 4.0.2a will shortly force a mid-cycle outage due to the 3.25 times restriction of Surveillance Requirement 4.0.2b.

The factors that could potentially affect CEA drop times will not adversely change as a result of extending the test interval from 18 months to every refueling outage. These factors include changes in component clearances, changes in the physical configuration of the CEA or guide tube, and the build-up of corrosion products and suspended material in the RCS that could interfere with CEA motion.

Changes in component clearances or the phsyical configuration of the CEA/.

, guide tubes are most likely to occur due to reactor head removal and replace-ment, or maintenance on the CEA and that portion of the drive system directly interfacing with the assembly. For these cases Surveillance Requirements 4.1.3.4a_and 4.1.3.4b, respectively, require CEA rod drop testing independent of the proposed change. Chemistry requirdments (e.g.' Technical Specification 3.4.6) and other controls on the reactor coolant system will minimize )

corrosion and the build-up of corrosion products or other suspended materials l

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_2-in those areas affecting CEA drop times. Additionally, each full length CEA is exercised every 31 days in accordance with Surveillance Requirement 4.1.3.1.2. This surveillance will detect rods that are sticking. And, the loose parts detection system (Technical Specification 3.3.3.9) will alert the operating staf f to conditions with a potential for af fecting reactor internals.

Although not documented in the Bases, the 18 month surveillance interval for rod drop testing was intended to accommodate a reactor refueling cycle.

The proposed change will explicitly identify that intent while retaining other Technical Specification controls to ensure rod drop testing for situa-tions that could affect rod drop times.

Approval of the proposed change is requested by December 1, 1986 in order to avoid a forced outage.

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: NO The proposed change will make explicit the intent to perform rod drop timing tests at refueling outages. Factors that could adversely affect rod drop times during a cycle (e.g. CEA drive maintenance) are addressed through other Surveillance Requirements not affected by the proposed change. Therefore the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: NO The proposed change only redefines the surveillance interval for rod drop testing. It introduces no physical modification to the plant nor does it implement a change in how the facility is operated. Therefore the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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3. Will operation of the facility in accordance with this proposed change l involve a signficant reduction in a margin of safety?

Response: NO l

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In redefining the rod drop testing surveillance interval to a refueling outage the original intent of the surveillance interval is preserved.

Actual interval periods may be somewhat longer or shorter than the nominal 18 months depending on plant availability for a particular cycle of operation. The actual factors affecting rod drop times continue to be controlled under other Technical Specifications. Therefore operation of Waterford 3 in accordance with the proposed change will not involve a significant reduction in a margin of safety.

Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3)

this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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NPF-38-20 ATTACHMENT A

REACTIVITY CONTROL SYSTEMS

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CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and regulating) CEA drop time, from a fully withdrawn position, shall be less than or equal to 3.0 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90% insertion position with:

a.

T,yg greater than or equal to 520*F, and

b. All reactor coolant pumps operating.

APPLICA8ILITY: MODES 1 and 2.

ACTION:

a. With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a. For all CEAs following each removal and reinstallation of the reactor vessel head,

b. For specifically affected individuals CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At least once per 18 months.

i s._I WATERFORD - UNIT 3 3/4 1-23

a NPF-38-20 ATTACHMENT B 4

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REACTIVITY CONTROL SYSTEMS

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@ CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and regulating) CEA drop time, from a fully withdrawn position, shall be less than or equal to 3.0 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90% insertion position with:

a. T,yg greater than or equal to 520 F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the CEA drop times within limits but determined at less than i

full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a. For all CEAs following each removal and reinstallation of the reactor vessel head,

b. For specifically affected individuals CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At lent :n;; p;r IC naths. E AcH /eEFi>EL/** O '

t WATERFORD - UNIT 3 3/4 1-23

NPF-38-21

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-21 The following is a correction related to the fuel rod loading value found in Technical Specification 5.3.1.

Existing Specification See Attachment A Proposed Specification See Attachment B Description In the present Technical Specification 5.3.1, the maximum fuel rod loading is 1807 grams of uranium. However, Combustion Engineering uses a nominal

( 426.55 KGU/ assembly loading (assuming 236 fuel red?, i.e. no burnable l poison pins) which is equivalent to the 1807 gram value.

The proposed change corrects this error by changing the phrase " maximum total weight" to " nominal weight".

Safety Analysis The proposed changes described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of i the following areas:  !

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1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed change is to correct an error in the technical specifications. It does not affect any of the safety analyses or the operation of the plant. Because of this, implementing this change will not increase the probability or consequences of any accident previously evaluated.

2. Will operation of the facility b acc rdance with the proposed change create the possibility of a ses er ferent kind of accident from any accident previously evaluateat Response: No.

Because no physical change is being made to the facility, and because there will not be a change in how the facility is operated, the proposed change will not introduce a new or different kind of accident from those previously : -;aated.

3. Will operation of this facility in accordance with this proposed change involve a significant reduction in margin of safety?

Response: No.

Because the proposed change is to correct an error in the technical specifications, it does not affect any of the safety analyses or the operation of the plant. Therefore, there is no impact on the margin of safety.

The Commission has provide guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48 CFR 14870) of amendments that are considered not likely to involve significant hazards considerations. Example (1) relates to a purely administrative change to the Technical Specifications:

for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature.

In this case, the proposed change is similar to Example (1) in that the change to Technical Specification 5.3.1 is a correction of an error.

Safety and Significant Hazards Determination Based upon the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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NPF-38-21 ATTACHMEfff A 1

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.- DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 236 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of 1807 grams uranium. The initial core loading shall have a maximum enrichment of 2.91 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.70 weight percent U-235.

CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 83 full-length and 8 part-length control element assemblies.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements,

b. For a pressure of 2500 psia, and

c. For a temperature of 650*F, except for the pressurizer and surge line which is 700 F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 11,800

+600, -0 cubic feet at a nominal T,yg of 582.1*F.

5.5 METEOROLOGICAL TOWERS LOCATION 5.5.1 The primary and backup meteorological towers shall be located as shown on Figure 5.1-1.

~ . .

WATERFORD - UNIT 3 5-5

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,ms y DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core'shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 236 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a 1404:NAL me*4mem total weight of 1807 grams uranium. The initial core loading shall l have a maximum enrichment of 2.91 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.70 weight percent U-235.

CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 83 full-length and 8 part-length control element assemblies.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 j of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements,

b. For a pressure of 2500 psia, and

c. For a temperature of 650 F, except for the pressurizer and surge line which is 700*F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 11,800

+600, -0 cubic feet at a nominal T,yg of 582.1*F.

5.5 METEOROLOGICAL TOWERS LOCATION 5.5.1 The primary and backup meteorological towers shall be located as shown on Figure 5.1-1.

WATERFORD - UNIT 3 5-5

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NPF-38-22 1

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-22 This is a request to revise Administrative Control 6.2.4.1, Shift Technical Advisor, and Table 6.2-1, Minimum Shif t Crew Composition.

Existing Specification See Attachment A Proposed Specification See Attachment B Description On October 28, 1985 the Commission published in the Federal Register (50 FR 43621) a policy statement on Engineering Expertise on Shif t. The policy statement (similar to SECY-85-150) provided two options to meet the NUREG 0737 requirement for an on-duty shift Technical Advisor (STA):

Option 1: Combined SR0/STA Position - This option allows an SRO to perform the dual SRO/STA function provided that the SRO meets the STA training criteria of NUREG 0731, Item I.A.1.1 and possesses one of the educational require-ments defined in the policy statement.

Option 2: Continued Use of STA Position - This option allows a dedicated STA on each shif t who meets the STA training criteria of NUREG 0737 Item I.A.1.1.

A utility may use Option 1 on some shif ts and Option 2 on other shif ts, or may use the same option on every shift. If Option 1 is used for a shift, then the separate STA position may be eliminated for that shift. As noted in the policy statement, the Commission prefers a combined SRO/STA position c.

(Option 1).

The proposed change implements the combined SR0/STA position while maintain-ing the flexibility to use Option 2 should Option 1 requirements not be met.

Specifically, Administrative Control 6.2.4.1 is rewritten to define the STA requirements consistent with the Commission policy statement. The proposed change for Table 6.2-1, which defines the minimum shif t crew composition, includes a footnote to implement the dual SR0/STA position provided the SRO meets the criteria defined in Administrative Control 6.2.4.1.

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a po.sitive finding in any of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: NO As described in the Supplementary Information (50 FR 43621) accompanying the policy statement, the dedicated STA position l was intended as an interim measure only until longer-term goals were achieved. The long-term initiatives - including shift staffing increases, training and qualification program improvements, hardware modifications, emphasis on human factors considerations, procedural upgrades, and development

. of extensive emergency response organizations - have collec-tively resulted in an improvement in the capabilities and l

qualifications of the shift crew and their ability to

! diagnose and respond to accidents. The long-term initiatives, which have been implemented for Waterford 3, offset any

- liability in combining the SRO/STA position. Therefore, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of 1 accident from any accident previously evaluated?

Response: NO The STA position is intended to provide engineering and accident assessment advice to the Shif t Supervisor in the event of abnormal or accident conditions. Because the STA function comes into play post-accident, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: N0 i

As previously noted, implementation of the long-term initiatives of NUREG 0737 offset any liability in combining the SR0/SIA position. Additionally, as stated by the Commission, experience has shown that an STA who is also an SRO is better accepted by the shift crew. The assessment and direction of an SR0/STA in an off-normal event might be better accepted by the crew than i.

1

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4 assessment and advice by a separate STA. Therefore, the i proposed change will not involve a significant reduction in a margin of safety.

The Commission has provided guidance concerning the application of i standards for determining whether a significant hazards consideration i

exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards considerations.

Example (vii) concerns a change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations. In this case, it is clear that the Commission intends licensees to implement a

{ dual SRO/STA position, and has determined that with the long-term NUREG 0737 improvements it is safety-beneficial to do so. While the Commission's policy statement is not, strictly speaking, a regulation, the proposed change is nonetheless similar to Example (vii).

Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that: (1) the proposed

change does not constitute a significant hazards consideration as defined j by 10CFR50.92; and (2) there is a reasonable assurance that the health and j safety of the public will not be endangered by the proposed change; and (3)

, this action will not result in a condition which significantly alters the l j impact of the station on the environment as described in the NRC Final i Environmental Statement.

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NPF-38-22 ATTACHMENT A

((. TABLE 6.2-1 V

MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, OR 4 MODE 5 OR 6 SS 1 1 SR0 1 None R0 2 1 A0 2 1 STA 1 None SS -

Shift Supervisor with a Senior Operator License SR0 -

Individual with a Senior Operator License R0 -

Individual with an Operator License A0 -

Auxiliary Operator STA -

Shift Technical Advisor Except for the Shift Supervisor, the shift crew composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shif t change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Supervisor from the control room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator or Operator license shall be designated to assume the control room command function.

WATERFORD - UNIT 3 6-5

ADMINISTRATIVE CONTROLS

)

4 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) l FUNCTION I 6.2.3.1 The ISEG shall function to examine unit operating characteristics, j NRC issuances, industry advisories, Licensee Event Reports, and other sources

of unit design and operating experience information, including units of similar design, which may indicate areas for improving unit safety. The ISEG

shall make detailed recommendations for revised procedures, equipment modifica-tions, maintenance activities, operations activities, or other means of improving j unit safety to the Engineering and Nuclear Safety Manager.

COMPOSITION 4 6.2.3.2 The ISEG shall .be composed of at least five, dedicated, full-time engineers located on site. Each shall have a bachelor's degree in i engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in

! the nuclear field.

RESPONSIBILITIES j 6.2.3.3 The ISEG shal.1 be responsible for maintaining surveillance of unit i activities to provide independent verification

• that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY l 6.2.3.4 The ISEG is an onsite independent technical review group that reports

, to the Engineering and Nuclear Safety Manager. The ISEG shall have the authority

necessary to perform the functions and responsibilities as delineated above.

1

! RECORDS i

! 6.2.3.5 Records of activities performed by the ISEG shall be prepared, maintained, and forwarded each calendar month to the Engineering and Nuclear Safety Manager.

l 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

j *Not responsible for sign-off function.

WATERt0RD - UNIT 3 6-6

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i NPF-38-22 ATTACHMENT B f

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TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, OR 4 MODE 5 OR 6 SS 1* 1 SR0 1S None M 2 1 A0 2 1 STA 1$ None l

SS -

Shift Supervisor with a Senior Operator License SRO -

Individual with a Senior Operator License RO -

Individual with an Operator License A0 -

Auxiliary Operator STA -

Shift Technical Advisor Except for the Shift Supervisor, the shift crew composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Supervisor from the control room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator or Operator license shall be designated to assume the control room command function.

$ An luowiou r toaru seo/srA Q UALIFICA MNS CAa Garisfy YHs s s/s t^A 0 9. S Ro lSM Posi now 2e0 umEM en rs 5/ MuLTANGOUS L Y.

WATERFORD - UNIT 3 6-5 f

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ADMINISTRATIVE CONTROLS a

6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of unit design and operating experience information, including units of similar design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifica-l' tions, maintenance activities, operations activities, or other means c.f improving unit safety to the Engineering and Nuclear Safety Manager.

COMPOSITION I

6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time l

engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level j

experience in his field, at least 1 year of which experience shall be in i

the nuclear field.

I RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification

• that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The ISEG is an onsite independent technical review group that reports to the Engineering and Nuclear Safety Manager. The ISEG shall have the authority necessary to perform the functions and responsibilities as delineated above.

RECORDS i

6.2.3.5 Records of activities performed by the ISEG shall be prepared, maintained, and forwarded each calendar month to tha Engineering and Nuclear Safety Manager.

! 6.2.4 SHIFT TECHNICAL ADVISOR I

j 6.2.4.1 Th Shift Techni Advisorshallp/videadvisory chnical support j to the S

• t Supervisor n the areas of the'rmal hydraulics reactor engineering, i and p1 t analysis w regard to the s e operation of e unit. The Shift Tec cal Advisor all have a bache r's degree or ivalent in a scientif l

- o engineering scipline and shal have received ecific training in th

/ response and alysis of the un for transient nd accidents, and in it design and yout, including e capabilities f instrumentation and ntrols j in the cpd rol room.

A Glase e r )

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WATERFORD - UNIT 3 6-6  :

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INSERT 6.2.4 Shift Technical Advisor 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shif t Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. The STA shall meet the requirements of either Option 1 or 2 as shown below

a. Option 1 - Combined SR0/STA Position. This

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option is satisfied by assigning an individual with the following qualifications to each operating shift crew as one of the SRO's required by 10CFR50.54(m) (2) (1):

1. Licensed as a Senior Reactor Operator on the unit and

2. Meets the STA Training Criteria of NUREG-0737, Item I.A.I.1, and one of the following educational alternatives:

(a). Bachelor's Degree in Engineering or Science from an accredited institution; (b) Professional Engineers License obtained by the successful completion of the PE examination; (c) Bachelor's Degree in Engineering or Science Technology from an accredited institution including course work in the physical, mathematical, or engineering sciences.

b. Option 2 - Dedicated STA Position. This option is satisfied by placing on each shift a dedicated Shift Technical Adviscr (STA) who meets the STA criteria of NUREG-0737, Item I. A.1.1.

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NPF-38-23

DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-23 This is a request to revise Technical Specification Table 3.3-1, Reactor Protective Instrumentation, to include the capability of bypassing the high steam generator level trip during Modes 1 and 2.

Existing Specification See Attachment A Proposed Specification See Attachment B Description In a continuing effort to reduce plant trips, LP&L will be implementing a station modification at the Waterford 3 first refueling outage to provide the plant operators with the capability of bypassing the high steam generator level reactor trip. The proposed change introduces the Technical Specifica-tion change needed to allow the operations staff to bypass the trip in Modes 1 and 2.

As noted in Section 2.2.1 of the Technical Specification Bases:

The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over.

Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over. This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the safety analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

The high steam generator level trip is described in Section 7.2.1.1.1.10 of the Waterford 3 FSAR. Because it is an equipment protective trip only, it does not fall within the scope of IEEE 279-1971, Criteria for Protection Systems for Nuclear Generating Stations. However, in order to enhance the overall reliability of the Reactor Protection System (RPS) and, as stated in the FSAR, "to preserve uniformity of function and design, the high steam generator level trip function meets the design bases" for other RPS components, including IEEE 279-71.

I The proposed change will not affect the design or testing of the non-safety related high steam generator level trip function but will only provide the option to bypass the function in Modes 1 and 2.

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Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a signficant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed change allows bypassing the non-safety related steam genera-tor level trip. This trip is not credited in the Waterford 3 safety analyses nor does the trip setpoint correspond to a Technical Specifica-tion safety limit. The design, testing and reliability of the RPS is unaffected by the proposed change. Therefore the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The most adverse consequence of bypassing the high steam generator level trip is the potential for moisture carryover to the turbine and subsequent turbine damage. This, however, is not a safety concern. The main steam line piping to the main steam isolation valves is designed to carry a water loading. Even should the main steam line piping be postulated to rupture due to the water loading, the resulting event is bounded by the main steam line break event analyzed in the FSAR. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No.

The high steam generator level trip is not credited in any safety evaluation. By definition, bypassing the trip cannot provide any reduction in a margin of safety.

Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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g TABLE 3.3-1 4

g REACTOR PROTECTIVE INSTRUMENTATION 2;

g MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

[ FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z

U 1. Manual Reactor Trip 2 sets of 2 1 set of 2 2 sets of 2 1, 2 1 w 1 set of 2 3*, 4*, 5*

2 sets of 2 2 sets of 2 8

2. Linear Power Level - High 4 2 3 1, 2 2#, 3#

3. Logarithmic Power Level-High

a. Startup and Operating 4 2(a)(d) 3 1, 2 2#, 3#

4 2 3 3*, 4*, 5* 8

b. Shutdown 4 0 2 3, 4, 5 4

4. Pressurizer Pressure - High 4 2 3 1, 2 2#, 3#

g 5. Pressurizer Pressure - Low 4 2(b) 3 1, 2 2#, 3#

[ 6. Containment Pressure - High 4 2 3 1, 2 2#, 3#

0 7. Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

8. Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

9. Local Power Density - High 4 2(c)(d) 3 1, 2 2#, 3#

10. DN8R - Low 4 2(c)(d) 3 1, 2 2#, 3#

11. Steam Generator Level - High 4/SG 2/SG 3/SG 1, 2 2#, 3#

12. Reactor Protection System Logic 4 2 3 1, 2 5 3*, 4*, 5* 8

13. Reactor Trip Breakers 4 2(f) 4 1, 2 5 3*, 4*, 5* 8

14. Core Protection Calculators 4 2(c)(d) 3 1, 2 2#, 3# and 7

15. CEA Calculators 2 1 2(e) 1, 2 6 and 7

16. Reactor Coolant Flow - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

r TABLE 3.3-1 (Continued)

TABLE NOTATION With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

The provisions of Specification 3.0.4 are not applicable.

~4 (a) Trip may be manually bypassed above 10 % of RATED THERMAL POWER; bypass shall be4 automatically removed when THERMAL POWER is less than or equal to 10 % of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 400 psia; bypass shall be auto- i matically removed whenever pressurizer pressure is greater than or equal to 500 psia.

~4 (c) Trip may be manually bypassed below 10 % of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10 % of4 RATED THERMAL POWER. During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 5%

of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 5% of RATED THERMAL POWER.

(d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

3

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(e) See Special Test Exception 3.10.2.

(f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.

ACTION STATEMENTS ACTION 1 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.

ACTION 2 -

With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6k. The channel shall be returned to OPERABLE status prior to STARTUP following the next COLD SHUTDOWN.

WATERFORD - UNIT 3 3/4 3-4

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NPF-38-23 ATTACllMENT B l

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3 TABLE 3.3-1 O

g REACTOR PROTECTIVE INSTRUMENTATION 2;

E MINIMUM

, TOTAL NO. CHANNELS CHANNELS APPLICABLE e FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 5

• 1. Manual Reactor Trip 2 sets of 2 1 set of 2 2 sets of 2 1, 2 1 w 2 sets of 2 1 set of 2 2 sets of 2 3*, 4*, 5* 8

2. Linear Power Level - High 4 2 3 1, 2 2#, 3#

3. Logarithmic Power Level-High

a. Startup and Operating 4 2(a)(d) 3 1, 2 2#, 3#

4 2 3 3*, 4*, 5* 8

b. Shutdown 4 0 2 3, 4, 5 4

4. Pressurizer Pressure - High 4 2 3 1, 2 2#, 3#

q 5. Pressurizer Pressure - Low 4 2(b) 3 1, 2 2#, 3#

[ 6. Containment Pressure - High 4 2 3 1, 2 2#, 3#

0 7. Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

8. Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

9. Local Power Density - High 4 2(c)(d) 3 1, 2 2#, 3#

10. DNBR - Low 4 2(c)(d) 3 1, 2 2#, 3#

11. Steam Generator Level - High 4/SG 3/SG 1, 2 2#, 3#

2/SG(3) l

12. Reactor Protection System Logic 4 2 3 1, 2 5 3*, 4*, 5* 8 l 13. Reactor Trip Breakers 4 2(f) 4 1, 2 5 l

3*, 4*, 5* 8

14. Core Protection Calculators 4 2(c)(d) 3 1, 2 2#, 3# and 7

15. CEA Calculators 2 1 2(e) 1, 2 6 and 7

16. Reactor Coolant Flow - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#

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TABLE 3.3-1 (Continued)

TABLE NOTATION With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

The provisions of Specification 3.0.4 are not applicable.

-4 (a) Trip may be manually bypassed above 10 % of RATED THERMAL POWER; bypass shall be4 automatically removed when THERMAL POWER is less than or equal to 10 % of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 400 psia; bypass shall be auto-matically removed whenever pressurizer pressure is greater than or equal to 500 psia. .

~4 (c) Trip may be manually bypassed below 10 % of RATED THERMAL POWER; bypass shall bg4 automatically removed when THERMAL POWER is greater than or equal to 10 % of RATED THERMAL POWER. During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 5%

of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 5% of RATED THERMAL POWER.

(d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

j (e) See Special Test Exception 3.10.2.

(f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.

> ) 1 ACTION STATEMENTS I ACTION 1 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.

ACTION 2 -

With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6k. The channel shall be returned to OPERABLE status prior to STARTUP following the next COLD SHUTDOWN.

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