RS-14-127, Response to Request for Additional Information Related to License Amendment Request to Revise Peak Calculated Primary Containment Internal Pressure

Response to Request for Additional Information Related to License Amendment Request to Revise Peak Calculated Primary Containment Internal Pressure

ML14163A690
Person / Time
Site:	LaSalle
Issue date:	06/12/2014
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-14-127
Download: ML14163A690 (11)

Text

RS-14-127 10 CFR 50.90 June 12, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

Response to Request for Additional Information Related to License Amendment Request to Revise Peak Calculated Primary Containment Internal Pressure

References:

1) Letter from D. M. Gullett (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "License Amendment Request to Revise Peak Calculated Primary Containment Internal Pressure," dated September 5, 2013

2) Letter from B. Purnell (U. S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Generation Company, LLC), "LaSalle County Station, Units 1 and 2 -

Request for Additional Information Regarding License Amendment Request to Revise Containment Peak Pressure (TAC Nos. MF2690 and MF2691),"

dated April 10, 2014 In Reference 1, Exelon Generation Company, LLC (EGC) submitted a license amendment request for LaSalle County Station (LSCS), Units 1 and 2, to revise the peak calculated primary containment internal pressure for the design basis loss-of-coolant accident (LOCA) described in TS 5.5.13, "Primary Containment Leakage Rate Testing Program." The peak calculated primary containment internal pressure, Pa. would be increased from 39.9 psig to 42.6 psig.

The proposed increase in Pa is the result of a LOCA Drywell Temperature sensitivity analysis performed by General Electric Hitachi (GEH).

In Reference 2, the U.S. Nuclear Regulatory Commission (NRC) requested additional information to complete its review of the proposed license amendment request. Attachments 1 and 2 to this letter provide the requested information. Reference 2 requested a response by May 28, 2014. As discussed with Mr. Blake Purnell (NRC) on May 23 and May 28, 2014, it was agreed that EGC would provide the response by June 12, 2014.

EGC has reviewed the information supporting a finding of no significant hazards consideration, and the environmental consideration, that were previously provided to the NRC in Attachment 1 of Reference 1. The additional information provided in this submittal does not affect the bases

June 12, 2014 U.S. Nuclear Regulatory Commission Page2 for concluding that the proposed license amendment request does not involve a significant hazards consideration. In addition, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Ms. Lisa A. Simpson at (630) 657-2815.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 12th day of June 2014.

Respectfully, David M. Gullatt Manager - Licensing Exelon Generation Company, LLC

Attachment:

Response to Request for Additional Information cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, LaSalle County Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT Response to Request for Additional Information By letter dated September 5, 2013, Exelon Generation Company, LLC (EGC) submitted a license amendment request (LAA) for LaSalle County Station (LSCS), Units 1 and 2, to revise the peak calculated primary containment internal pressure for the design basis loss-of-coolant accident (LOCA) described in TS 5.5.13, "Primary Containment Leakage Rate Testing Program," (Reference 1). The peak calculated primary containment internal pressure, Pa, would be increased from 39.9 psig to 42.6 psig. In a letter dated April 10, 2014, the NRC requested additional information to complete its review of the proposed LAA (Reference 2).

NRC RAl-1:

Describe the containment models used in the revised analysis for mass and energy release and containment response. The response should identify the models used: (1) describe the differences, if any, between these models and those in the analysis of record (AOR) and (2) identify if the models are NRG-approved. Clarify whether the analyses were performed for short-term only or for both short-term and long-term, and if both loss-of-coolant accident and main steamline break were analyzed.

EGC Response to NRC RAl-1:

The Analysis of Record (AOR) for the LSCS containment system is GE Project Task Report GE-NE-A1300384-02-01-R3, which is documented in Design Analysis L-002874 (Reference 3).

The proposed increase in Pa is the result of a LOCA-Drywell Temperature sensitivity analysis performed by General Electric Hitachi (GEH). EGC requested GEH to perform the sensitivity analysis in support of the LSCS Extended Power Uprate (EPU) project that was subsequently canceled. The GEH sensitivity analysis supporting the EGC LAR submitted September 5, 2013 (Reference 1) is GEH 0000-0149-2311-RO, Revision O (2012 GEH Analysis, Reference 4).

In the AOR, two separate analyses are performed:

• Short Term Containment Pressure Response This analysis determines the short term pressure and temperature responses of the Containment (Drywell and Wetwell) to postulated LOCAs. The short term evaluation addresses the time period of O to 20 seconds. This analysis determines the peak containment (Drywell) pressure, which occurs at approximately 12 seconds following the LOCA. It also determines the maximum Drywell-to-Wetwell pressure differential, which occurs within the first five seconds.

• Long Term Containment Pressure Response This analysis determines the long term pressure and temperature responses to determine the peak suppression pool temperature. The long term evaluation addresses the time period from Oto approximately 100,000 seconds (28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />) following the design basis accident (DBA) LOCA.

Only the short term response was included in the 2012 GEH Analysis (Reference 4).

Containment long term analysis was not included in the LAA as the short term analysis determines the peak containment (Drywell) pressure and the Drywell-to-Wetwell pressure difference.

Page 1of9

ATTACHMENT Response to Request for Additional Information The 2012 GEH Analysis only considered the LOCA; it did not consider the main steamline break. As described in UFSAR 6.2.1.1.3.1, the limiting event is the instantaneous guillotine rupture of a recirculation line. This break results in the maximum flow rate of primary system fluid and energy into the Drywell.

As explained in the 2012 GEH Analysis (Reference 4), the models/methodology used in the analysis performed to support the LAR are the same as that used in the short term response portion of the AOR that determines the peak containment pressure. This is summarized in the following table:

Short Term Response to the OBA LOCA Analysis of Record 2012 GEH Analysis Purpose (GE-NE-A 1300384-02-01-R3, (GEH 0000-0149-2311-RO, dated May 2002) dated August 2012)

Computer code used to calculate the LAMB-08 LAMB-08 mass & energy released from the break Computer code used to determine the short term pressure and temperature M3CPT05V M3CPT05 responses of the Containment (Drywell (See Note 1) (See Note 1) and Wetwell) to postulated LOCAs.

Computer code used to determine Wetwell airspace pressurization produced by Wetwell airspace N/A PICSM01V compression with pool swell for (See Note 2) determination of peak Drywell-to-Wetwell pressure difference.

Note 1: The computer codes are the same but are run on different environments. M3CPT05V is run on a VAX computer. M3CPT05 is run on an "ALPHA System" computer.

Note 2: Peak values of Drywell-to-Wetwell pressure difference that are reported in the AOR include values that are based on M3CPT05 results directly and also values that account for Wetwell pressurization due to Wetwell airspace compression during pool swell as determined with the PICSM code. The values for peak Drywell-to-wetwell pressures reported in the 2012 GEH Analysis were based on results obtained directly from the M3CPT code.

The use of the M3CPT code has been accepted by the NRC for calculating the short-term response of the containment system to LOCAs from the start of the transient until operator intervention via the Automatic Depressurization System (ADS) or until the reactor blowdown is complete, whichever comes first. The GE containment analysis methods have been reviewed by the NRC (References 5 and 6). The use of LAMB-08 blowdown flow in M3CPT was identified in Appendix G of NEDC-31897P-A, May 1992, which was approved by the NRC.

The use of the PICSM code for calculation of the pool swell response, including Wetwell airspace pressurization with pool swell, was approved by the NRC in NUREG-0808 (Reference 5).

Page 2 of 9

ATTACHMENT Response to Request for Additional Information NRC RAl-2:

Describe any other input changes to the containment response models between the AOR and the revised analysis which are not already identified in the LAA.

EGC Response to NRC RAl-2:

There are two types of changes regarding the LSCS LAA analysis against the LSCS current AOR. The first type is the analysis input change and the second type is containment analysis model option selection. The input changes have been discussed in the LAA and are listed for reference as follows:

• Initial Drywell Temperature

• Main Steam Safety Relief Valve (SRV) Setpoints

• Main Steam Isolation Valve (MSIV) Closure Time

• Core Flow Conditions Permitted by the Power/Flow Map

• Reactor power level Additionally, two containment analysis model options were updated based upon the current GE Hitachi containment analysis methodology. Two LAMB model options have been used differently from LSCS AOR: the critical break flow option and the option for core heat flux (representing core decay heat, fission power, and fuel relaxation energy). Two versions of Moody Slip model are available in LAMB for the calculation of critical break flow. The AOR used the LAMB option implementing the mechanistic slip flow model for subcooled critical flow, whereas the current LAA uses the option which extrapolates the subcooled flow calculation based on the saturated-to-subcooled fluid density ratio. Regarding the core heat flux, there are two options (curves) available in LAMB model to accommodate fuel design changes. Both of them were derived conservatively for LAMB applications. The LAA analysis has used the bounding curve of these two curves. It is further determined that these two options selected for the LAA analysis have insignificant impact on the LSCS LAA containment analysis, and their implementation retains the overall conservative behavior of the GE Hitachi containment analysis methodology.

NRC RAl-3:

The LAA provides the contribution to the increased Pa from initial drywell temperature as 1.3 pounds per square inch gauge (psig), and from the secondary issues as 1.4 psig. However, the LAA does not describe how the Pa increase from the change in initial drywell temperature was determined. The AOR used a thermal power of 3789 megawatts-thermal (MWt) and the revised analysis used a thermal power of 3559 MWt, but the LAA does not state which thermal power was used to determine the increase in Pa from the initial drywell temperature.

Describe how the contribution to the Pa increase from the change in initial drywell temperature was determined.

Page 3 of 9

ATTACHMENT Response to Request for Additional Information EGC Response to NRC RAl-3:

The containment AOR (Reference 3) was performed at a power level of 3789 MWt and not at the 3559 MWt power level identified in the AOR. As a consequence, the peak calculated primary containment internal pressure, Pa, determined in the containment AOR is conservative; it is approximately 0.5 psi higher than it would be at 3559 MWt.

The current containment AOR (Reference 3) uses the maximum allowed Drywell temperature permitted by the TS 3.6.1.5 during power operations (i.e., 135 °F). The increased Pa of 1.3 psig as the result of using lower initial Drywell temperature was evaluated as part of the 2012 GEH Analysis (Reference 4). The 2012 GEH Analysis was performed at a power level of 3559 MWt, which is 102 % of the stretch power uprate licensed thermal power (i.e., 102 % of 3489 MWt =

3559 MWt).

As identified in the following table, the 2012 GEH Analysis shows that use of the maximum allowed Drywell temperature may not be conservative in relation to the peak LOCA containment (Drywell) pressure. A lower Drywell temperature results in a 3 % higher peak containment pressure.

Drywell Temp Peak Drywell Power when LOCA Occurs Pressure 3559 MWt 135 °F 41.3 psig 3559 MWt 98 °F 42.6 psig Difference (Increase) 1.3 psiq (3 %)

As discussed in the EGC LAR submitted September 5, 2013, additional information from GEH revealed additional issues that also affected the peak calculated primary containment internal pressure.

NRC RAl-4:

The LAR states that containment leak rate testing based on the higher Pa would increase the measured leakage, and that "EGC Operability Evaluation OE 11-002 demonstrates that there is adequate margin to accommodate this increase." However, the LAR does not provide information to support this statement. The NRC staff needs the following information to reach a conclusion regarding the adequacy of containment leak tightness:

A. Provide the last two integrated leak rate test results for each unit, as a percentage of the maximum allowable primary leakage rate, and the test pressures.

B. Provide the latest combined Types B and C test summation at the new Pa. Identify when all these tests are scheduled to be completed at the new calculated maximum accident pressure. If all past tests were done at a pressure exceeding the new calculated Pa, a simple statement to that effect would suffice.

C. Provide the margins available to the leakage rate acceptance criteria (Types A, B, and C),

for both the current Pa and the new Pa. Explain how it was determined that there is adequate margin to accommodate the Pa increase (e.g., identify extrapolation methods).

Page 4 of 9

ATTACHMENT Response to Request for Additional Information EGC Response to NRC RAl-4.A:

The following table summarizes the last two integrated leak rate test results for LSCS, Units 1 and 2. The table compares the measured and allowable leakage rates, identifies the margin, and identifies how the higher Pa (i.e., 42.6 psig) would affect the measured leakage.

lnte ;irated Leak Rate Testing Allowable Impact of Higher Pa Min Test Measured When Performed Leakage Margin (42.6 psig) upon Pressure Leakage (See Note 3) Measured Leakage 26%

42.483 psig 2008 (L 1R 12) 0.472 % I day 0.635 % I day 0.14 % increase (0.163 % I day)

Unit 1 58% 2.56 % increase 40.5 psig 1994 (L 1 R06) 0.266 % I day 0.635 % I day (0.369 % I day) <See Note 4) 39%

41.2 psig 2009 (L2R12) 0.386 % I day 0.635 % I day 1.68 % increase (0.249% I day)

Unit 2 33% 1.81 % increase 41.1 psig 1993 (L2R05) 0.427 % I day 0.635 % I day (0.208 % I day) (See Note 4)

Note 3: When the previous two Integrated Leak Rate Tests for each unit were performed, the allowable leakage (La) was 0.635% per day. The implementation of Alternative Source Term (AST) at LSCS by NRC letter dated September 6, 201 O (Reference 7) increased the allowable leakage from 0.635% to 1% per day.

Note 4: The impact of the higher Pa upon the measured leakage is provided for information only. These tests were performed prior to power uprates. Therefore, the Pa applicable when the tests were performed would be< 42.6 psig.

EGC Response to NRC RAl-4.B:

LSCS Unit 1 recently completed refueling outage L 1R15. The local leak rate tests (LLRT) from L 1R15 were performed at a test pressure of at least 42.6 psig. Preliminary assessment indicates that significant margin exists. The formal report evaluating the results is in preparation stage.

Since the formal results are not yet available, the following table summarizes LLRT test results performed during previous refueling outages (L 1R14 and L2R14).

Type B & C Local Leak Rate Test Results Min Test Allowable When Performed As-Left Margin Pressure (Reference 8)

Min Path 101.23 scfh 263.17 scfh (72 %)

Unit 1 40.85 psig L1R14(2012) 364.4 scfh Max Path 206.39 scfh 158.01 scfh (43 %)

Min Path 73.38 scfh 291.02 scfh (80 %)

Unit 2 40.61 psig L2R14 (2013) 364.4 scfh Max Path 160.93 scfh 203.47 scfh (56 %)

Only a portion of the primary containment isolation valves (PCIVs) are tested in any one outage.

Therefore, it takes several outages for the full complement of PCIVs (i.e., all 167 Unit 1 and 146 Unit 2 primary containment penetrations) to be tested. The schedule for performing these tests is in accordance with the LSCS Appendix J program.

Page 5 of 9

ATTACHMENT Response to Request for Additional Information EGC Response to NRC RAl-4.C:

The available margins for the Type A integrated leak rate test results and the effect of the higher Pa upon the margin was discussed in the response to NRG RAl-4.A. The available margins for the Type B & C LLRT results and the effect of the higher Pa upon the margin were discussed in the response to NRG RAl-4.B.

The higher Pa (i.e., 42.6 psig) will affect the containment leak rate testing results. This effect will be quantified by modeling the leakage as compressible flow through an orifice. From Crane's Technical Paper No. 41 O (Reference 9), the compressible fluid through an orifice in standard cubic feet per hour (scfh) is as follows:

Where:

q'h =flow in scfh Y = net expansion factor for compressible flow through orifices d1 = pipe internal diameter C = flow coefficient for orifices Sg = specific gravity of the gas relative to air

~p = differential pressure across the orifice P1 = weight density of the fluid Since the parameter affected is the differential pressure ~P (See Note 5):

Therefore, as shown below, increasing Pa from 39.9 psig to 42.6 psig, the expected leakage rate would result in a 3.3 % increase in the measured leakage rate.

q'hb I q'ha = (42.6 psig) 0*5/(39.9 psig) 0*5 = 1.033 q'hb = (1.033)q'ha This would result in a 3.3 % increase in the measured leakage rate.

For the Unit 1 integrated leakage rate test performed at 42.483 psig, the expected increase in leakage rate would be 0.14 %.

q'hb I q'ha = (42.6 psig)°-5/(42.483 psig) 0 *5 = 1.0014 q'hb = (1.0014)q'ha For the Unit 2 integrated leakage rate test performed at 41.2 psig, the expected increase in leakage rate would be 1.68 %.

q'hb I q'ha = (42.6 psig)°-5/(41.2 psig)°" 5 = 1.0168 q'hb = (1.0168)q'ha Note 5: It is recognized that other parameters are pressure dependent, but because the pressure changes are small, these effects are considered insignificant.

Page 6 of 9

ATTACHMENT Response to Request for Additional Information NRC RAl-5:

Technical specification (TS} limits on drywall air temperature are necessary because the parameter satisfies Criterion 2 of Title 10 of the Code of Federal Regulations (10 CFR),

Section 50.36(c}(2)(ii). Specifically, the drywall air temperature is a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Traditionally, the maximum temperature in the drywall is included in the TS to ensure that the Pa value for containment and the environmental qualification temperatures for equipment located in the drywell are not exceeded during design-basis accidents. Since the LAR proposes to incorporate the effect of a lower initial containment temperature on Pa, a minimum drywell temperature is required in the TSs. The LAR did not include any proposed TS changes in this regard.

Provide proposed TS changes to add a minimum drywall temperature requirement, and provide a summary statement of the bases or reasons for such specification. Alternatively, explain why a minimum drywell temperature TS requirement is not necessary tor the TS to meet the requirements of 10 CFR 50.36, "Technical Specifications."

EGC Response to NRC RAl-5:

EGG has reviewed the information presented in this RAI, 10 CFR 50.36, the NRG Final Policy Statement on TS Improvements, and the LSCS UFSAR and has concluded that a new TS for a minimum drywell temperature is not necessary. This conclusion is based on the following considerations:

(1) The stated purpose of the TS limit on the maximum drywall air temperature is to control the initial condition to ensure that the peak LOCA drywell temperature does not exceed the maximum allowable temperature of 340 °F.

(2) The existing drywell air temperature TS provides the upper bounds for the range of values to be used in the accident analysis.

(3) Operational Experience does not support the addition of the new TS.

10 CFR 50.36(c)(2)(ii) states that a technical specification limiting condition for operation must be established for each item meeting one or more of four criteria. Criterion 2 of this section states:

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The NRG Final Policy Statement on TS Improvements discusses the content of TS and provides background on each of the four criteria of 10 CFR 50.36(c)(2)(ii). Specific to Criterion 2, the statement says that the plant shall be operated within the bounds of the initial conditions assumed in the existing accident analysis. The process variables or operating restrictions used in Criterion 2 are only those parameters for which specific values or ranges of values have been Page 7 of 9

ATTACHMENT Response to Request for Additional Information chosen as reference bounds in the accident analysis and are controlled during power operation such that the value remains within the analysis bounds.

As stated in the TS Bases for the drywell air temperature TS (i.e., TS Bases 3.6.1.5), with an initial drywell temperature less than or equal to the LCO temperature limit, the resultant accident temperature profile assures that the drywell structural temperature is maintained below its design limit and safety related equipment will continue to perform its function.

Specifically for LSCS, maintaining the expected drywell air temperature less than or equal to a specified value ensures that safety analyses remain valid and ensures that the peak post-accident drywell temperature does not exceed the maximum allowable temperature of 340 °F.

Exceeding this design temperature of 340 °F may result in the degradation of the primary containment structure under accident loads. Equipment inside primary containment that is required to mitigate the effects of an accident is designed to operate and be capable of operating under environmental conditions expected for the accident.

Similarly, the TS Bases for the drywell pressure TS (i.e., TS Bases 3.6.1.4) states that the LCO value preserves the initial conditions assumed in the accident analysis. This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions to a value below the TS pressure value. With the drywell pressure below this initial value, the resultant peak drywell accident pressure will be maintained below the design pressure.

Therefore, the TS for drywell air temperature and drywell pressure are established to limit conditions during power operations to ensure that the initial conditions in the accident analysis are not exceeded. This applies to parameters that must be controlled during power operations to ensure that the values used in the accident analysis remain within the analysis bounds. The 98 °F used in the analysis was based upon historical studies of the bulk average Drywell temperature and was selected to ensure that it is bounding. Therefore, a limit on the minimum Drywell temperature is not required. This is evidenced by the following data:

• The bulk average Drywell temperature is maintained > 100 °F when the reactor power is ;:;:: 90 % CLTP. The containment analysis used a lower initial Drywell temperature (i.e., 98 °F) to provide conservatism.

• A review of historical data indicates that the bulk average Drywell temperature is less than 100 °F at low power levels (i.e.,< 40 % CLTP).

As demonstrated in the analysis supporting the September 2013 LAA (2012 GEH Analysis, Reference 4), drywell air temperature is a parameter in the peak drywell pressure calculation.

With regard to impact on calculating peak drywell pressure, the TS drywell air temperature value is considered an upper bound for the maximum value that may be used in the accident analysis when calculating peak drywell pressure. Consistent with the NRG Final Policy Statement on Criterion 2, the high drywell temperature TS limit is considered the operating restriction for the range of values that can be used in the peak drywell pressure calculation. This range is the reasonable set of drywell air temperature values, below the TS limit, that must be considered in determination of post-accident peak pressure. In order to determine a bounding post-accident pressure, the Pa calculation should use a conservative value within the range controlled by the existing temperature TS. By selecting the initial drywell air temperature in the Pa calculation at or below the TS value, the resulting Pa value continues to maintain a limiting known margin to the design limit.

Page 8 of 9

ATTACHMENT Response to Request for Additional Information REFERENCES

1) Letter from D. M. Gullott (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "License Amendment Request to Revise Peak Calculated Primary Containment Internal Pressure," dated September 5, 2013

2) Letter from B. Purnell (U. S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Generation Company, LLC), "LaSalle County Station, Units 1 and 2 - Request for Additional Information Regarding License Amendment Request to Revise Containment Peak Pressure {TAC Nos. MF2690 and MF2691)," dated April 10, 2014

3) GE Project Task Report GE-NE-A1300384-02-01-R3, Revision 3, "LaSalle County Station Power Uprate Project Task 400: Containment System," dated May 2002 (documented in Design Analysis L-002874, Revision O)

4) GE Hitachi Nuclear Energy Analysis 0000-0149-2311-RO, Revision O, "LaSalle County Generating Station Units 1 & 2, Short Term Containment Bounding Pa Assessment,"

dated August 2012

5) NRC NUREG-0808, "Mark II Containment Program Load Evaluation and Acceptance Criteria," dated August 1981

6) NRC NUREG-0800, Standard Review Plan, Section 6.2.1.1.C, "Pressure-Suppression Type BWR Containment," Revision 6, August 1984

7) Letter from C. Gratton (U. S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Generation Company, LLC), "LaSalle County Station, Units 1 and 2 - Issuance of Amendments Re: Application of Alternative Source Term {TAC Nos. ME0068 and MF0069)," dated September 6, 2010

8) LaSalle Station Technical Surveillance LTS-300-5, "Primary Containment Leak Rate Testing Program," Revision 41, February 7, 2014

9) Crane Technical Paper No. 410, "Flow of Fluids through Valves, Fittings, and Pipe," 25th printing, King of Prussia, PA.: Crane Co., 1988 Page 9 of 9