ML19343C653

Safety Evaluation Supporting Amend 23 to License DPR-34

ML19343C653
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Site:	Fort Saint Vrain
Issue date:	03/16/1981
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8103240847
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SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT 23 TO FACILITY OPERATING LICENSE NO. DPR-34 0F PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 1.0 Introduction Fort St. Vrain, 330 MWe high temperature gas-cooled reactor (HTGR),

was designed by the General Atomic Company (GAC) and is operated by the Public Service Company of Colorado (PSCo) near Platteville, Colorado. PSCo was issued a construction permit on September 17, 1968 and submitted the Final Safety Analysis Report as Amendment 14 to its application for a construction pemit and operating license for the Fort St. Vrain Nuclear Generating Station (FSV) on November 4, 1969.

A Safety Evaluation Report dated January 20, 1972 and a first supplement which was issued on June 12, 1973 concluded that FSV can be operated, as proposed, at power levels up to 842 MWt, full 100 percent power, without endangering the health and safety of the public.

Af ter issuing the 1972 SER, several deficiencies were found which, in later years, limited the power level at which Fort St. Vrain could be operated. These were:

1.

In addition to the restrictions imposed by the Technical Specifica-tions on Fuel Loading and Initial Rise to Power, a 2 percent hold was imposed on FSV due to a cracked Pelton wheel by Commission letter dated November 14, 1974. The "A" helium circulator had been removed from its nomal position in the PCRV bottom head penetration and replaced with a spare due to problems with the static seal bellows.

Removal of the circulator made possible a detailed examination of the circulator components.

A fluorescent penetrant examination of the Pelton wheel compling area indicated uniform cracking at the root of all the coupling

. teeth and revealed six out of twenty pelton turbine buckets with cracks at-the root of the splitter. A commitment by PSCo to replace the cast pelton. wheel with a forged pelton wheel and to decrease the speed of the pelton wheel from 10,500 rpm to 7,000 rpm resulted in a SER granting continuation of rise-to-power testing above the previously imposed 2 percent limit.

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J. 2.

In April 1975, shortly after the occurrence of an electrical cable fire at the Browns Ferry plant, an inspection revealed that some fire stops in the electrical cable system had not been installed and that routing of some cables deviated from the installation criteria set forth in the FSAR.

By letter dated June 17, 1975 PSCo stated that FSV would be maintained in a subcritical condition, a condition it had been at since May 1,1975, until the problem in resolved to the satisfaction of the NRC.

The scope of consideration was broadened beyond electric cable segregation and separation to include fire prevention, detection and suppression, and alternate methods for accomplishing orderly plant shutdown and cooldown in case of loss of normal and preferred systems.

On June 18, 1976 the Commission issued Amendment No.14 to the license approving all proposed corrective actions and operation of FSV up to a power level of 40 percent of rated power. The Commission determined that "upon completion of all Stage 1 corrective actions, the FSV plant will achieve the same safety objective and provide the same margins of safety that were previously found acceptable for full power operation".

3.

During Phase.1 operations, testing and operational experience have indicated that certain changes in the plant should be made to improve operational reliability and safety; it is unlikely that these items would have been discovered without such experience.

In additidn, PSCo and General Atomic had identified, by letter dated March 1,1977, an inconsistency between the Facility Technical Specifications and the FSAR, performed accident analyses related to this matter, and have indicated that they be limited to 70 percent of rated power to remain within the conditions des'cribed in the FSAR until the matter is further resolved. Three items were subsequently identified by letter to PSCo as constituting a 70 percent

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hold on further reactor operation:

(a) accident reanalysis using correct power-to-flow ratios, (b) moisture injection tests and response times, and (c) time available before depressurization is necessary following LOFC.

These three items were discovered to be inconsistent with the FSAR and were addressed in' meetings and requests for revisions to the Technical Specifications. The staff reviewed the proposed revisions and concluded that they constitute corrections that result in data and analyses consistent with the FSAR.

Therefore, all requirements for continued rise-to-power have been satisfied.

The operating license, OPR-34, was issued on December 21, 1973 and has been amended twenty-three times including the amendment supported by this safety evaluation. A listing and brief description of the twenty-two prior amendments is-presented in Appendix A.

e s.

This amendment revises the Technical Specifications to: (1) extend the minimum sample flow limits to cover the reactor power range of 70 to 100

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percent, (2) define the times to depressurization, (3) extend the core residence time of the fuel test elements and (4) specify operator action time limits for power-to-flow ratios as per S.L. 3.1.

The reactor achieved criticality'on January 31, 1974, and low power physics testing was initiated. These low power tests, identified as the "A Series" tests, along with the "B Series," or pcwer ascension.. tests were reported in accordance with Section 7 of the Technical Specifications.

Also, in accordance with the Technical Specification, Public Service of Colorado provides " Reportable Event" reports and " Unusual Event" reports on safety items relating to abnormal, unusual or unanticipated events that occur during the course of plant operations.

In 4ddition to the reports received from the licensee, our safety reviews aave bene-fitted from information on plant status and operations provided by the Office of Inspection and Enforcement, and by visits to the plant site

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by technical specialists to review plant records and the "as-built" ccadition of the plant. Our safety review has also. included consideration of comparable light water reactor safety under the sponsorship of the Office of Nuclear Regulatory Research, and information developed during the review of the General Atomic Standard Safety Analysis Report, GASSAR.

2.0 Core Residence Time of Fuel Test Elements

2.1 Background

Eight fuel test elements (FTE's) were installed in the Fort St. Vrain (FSV) reactor core during the first refueling.

Approval for installation and operation of the eight FTE's had been granted via a letter (reference 1),

dated April 20, 1979, in which the NRC incorporated data from a tabulated list of design features that PSC had earlier subnitted as part of the safety analysis report (reference 2) for the proposed fuel test elements.

That tabulation (Technical Specification Table 6.1-1) included a list of residence. times for the FTE's.-

The FTE resident times in the as-submitted, original (reference 2a), Table 6.1-1 corresponded to the then-expected Fort St. Vrain reactor core operating and refueling schedule.

At that time, if the plant were to be operating at rated full power conditions, the first segment of the core was' to be replaced after six months of operation;~i.e.,150 equivalent full power days (where one equivalent ^ full power year is defined as 300 The second segment was then to be replaced after the full power days).

second six months (an additional 150 equivalent full power days), and subsequent refuelings were to be performed annually.. The reason for the short first two cycles was that the first two segments of the core contained fuel particles that were sligh'tly out of specification, thus, out of concern over fuel particle integrity, the fuel vendor (General Atomic) and the licensee agreed to reduce the projected residence time

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_4 for the first two fuel segments by half (reference 3).

This change is reflected in Amendment No. 24 to the Fort St. Vrain FSAR (refenence 4).

In PSC's view, the original basis for the reduction in length of the first two refueling cycles has now "been found to be without merit" (reference 3),

essentially because (1) FSV operating experience has shown that the fuel particles have performed very satisfactorily (as indicated by the contained low levels of primary' coolant activity), and (2) PSC has analyzed the potential affects of extending the first two refueling cycles from 150 equivalent full power days (EFPDs) to 200 EFPDs (reference 5), and has concluded (reference 6) that no unreviewed safety question exists. PSC, therefore, believes that no changes to the Technical Specifications are necessary with regard to the residence time limit for standard (i.e.,

reference) fuel. However, because Technical Specification Table 6.1-1 addresses, in part, the residence time of the eight FTE's currently in the core, and because the current approved version of the table lists the residence time of FTE-1 as-0.5 years (with corresponding cumnulative times for FTE-2 through FTE-8), a technical specification change is required.

The change would allow the extension of the residence times for the FTE's (and indirectly for the standard fuel elements as well, since all the reference fuel in a given segment is replaced at the same refueling outage).

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2.2 Staff Evaluation PSC's safety analysis conducted on the proposed extension of the FTE residence time was submitted as " Amendment 3 to the Safety Analysis Report for Fort St. Vrain Reload 1 Test Elements FTE-1 through FTE-8, GLP-5494,"

(reference 7). Our 1978/79 evaluation (reference 1) of the FTE safety analysis report (SAR), however, did not address the residence time extension in Amendment 3, and thus did not approve the extension of Cycle 2 operation of the FTE's. As indicated in our original evaluation of the FTE SAR, we believe that-the SAR (GLP-5494) demonstrated analytically that the inclusion of the test elements in the FSV reactor should improve the stress margins relative to the reference core and that the reduction in graphite dimensional changes should reduce column tilting _and resulting interfacial coolant flow.

Furthermore, results of the thermal analysis indicated that the FTE in-pile operating temperatures would be either essentially equal to, or less than

.the reference fuel elements being. replaced, with the exception of FTE-1.

That test element would operate at a higher temperature than the element it replaces, but only through the first cycle (approximately 6 months),

where it would be replaced by a standard element.

And fission product release calculations had indicated that gaseous fission product release

higher (which was not expected), while the effect on metallic fission product release was projected to be imperceptible.

Since the FSV primary coolant fission product inventory continues to be well below the original projected level (reference 3), there is no reason to believe that the rationale for the above conclusions has changed or is suspect.

Extending.the test element core residence time from 150 to 200 EFPD in the cycle 2 core thus does not alter the status of the improved safety margins resulting from insertion of and operation with the 8 FTE's, as discussed in GLP-5494.

2.3 Staff Position We have re-examined the bases for our original approval for insertion of the eight FTE's and have determined that extending Cycle 2 operation by 50 EFPDs does not alter these bases. Therefore, the status of the improved safe'ty margins resulting from insertion of, and operation with, the eight FTE's, as discussed in GLP-5494, remains the same. Thus,.he requested change neither imposes new significant hazards considerations nor alters our conclusion that adequate assurance has been presented that the facility can be operated without endangering the public health and safety.

2.4 References 1.

T. P. Spels (NRC), letter to C. K. Miller (PSC), April 20, 1979.

2.

C. K. Miller (PSC), letter to Richard P. Denise (NRC), January 9, 1978 (with attachments).

(a) Attachment 1, " Proposed Revision to Technical Specification 6.1 Reactor Core Design Features."

(b) Attachment 2, " Safety Analysis Report for Fort St. Vrain Reload 1 Test Elements FTE-1 through FTE-8," General Atomic Report GLP-5494, June 1977.

3.

Frederic E. Swart (PSC), letter (with attachments) to James F. Miller (NRC), January 21, 1981.

4.

Section 3.5.1. Nuclear Design Summary, Fort St. Vrain Final Safety Analysis Report, Amendment 24, p. 3.5-1.

5.

O. R. Lee (PSC), letter (with attachments) to James R. Miller (NRC),

October 3,1980.

6.

" Safety Analysis Report for Fuel Reload 1 Extended Operation, Fort St. Vrain Nuclear Generating Station, General Atomic Company Report

.GLP-5646,_ June 1, 1978.

7.

J. K.-Fuller (PSC), letter to W. P. Gammill (NRC), January 26, 1979, with attachment, Amendment 3 to the " Safety Analysis Report for Fort St. Vrain Reload 1 Test Elements FTE-1 through FTE-8," GLP-5494, June 30,-1977.

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h 3.0 Power-to-Flow Ratio Limits as per S.L. 3.1 3.1 Introduction On November 1,1977, the Public Service Company of Colorado submitted analyses in support of operation of the Fort St. Vrain plant at 100%

of design power.

The power level of the Fort St. Vrain plant was originally limited to 70% of design power because of limitations in the helium purification system which must be used for depressurization in the event of a loss of forced circulation accident. PSCo justified, through analysis, that at a power level of 70% of design power, temperature predictions would fall. at or below the original FSAR values.

It was during these reanalyses that discrepancies between the values for core region peaking factors and outlet tenperature dispersion used in the FSAR safety analyses and the values used in the plant technical specifications (which were higher) were identified.

Accident reanalyses using the more limiting initial operating conditions permitted by the technical specifications were then submitted in support of proposed full power operation for Ft. St.

Vrain.

Amendment No. 22 to the Fort St. Vrain license presented a Safety Evaluation Report dealing with accident reanalysis including connients on the Reactor Core Safety Limit dealing with power to flow ratios.

The applicant had stated that actual power to flow ratios may be in excess of 1.0 at indicated full power.

In addition, plant technical specifications permit operation at power-to-flow ratios in excess of 1.05 at power levels below 100% (see: Technical Specifications Figure 3.1-2).

Maximum temperatures however, would occur for the -105 percent power level case due to the increased decay heat generation.

This was confirmed by independent calculations by ORNL at selected points along the power-to-flow operating limit curve.

The Amendment 22 SER noted that in our review of initial conditions with respect to the allowable power-to-flow ratios in the technical specifications for limited periods of time, the technical specifi-

_ stions allow full power operation at power-to-flow ratios greater than 1.05 based on steady state time-at-temperature limits for fuel damage. The licensee considers operation in this region to be a degraded plant condition and has stated that normal practice is not to operate with power-to-flow ratios greater than 1.05.

Since operation in this degraded mode has not-been considered in the accident reanalyses, deliberate. operation at -power-to-flow ratios in excess of the curve shown.as Figure 3.1-1 of the technical specifications in not acceptable to the staff.

If the power-to-flow ratio limits of Figure 3.1-1. are exceeded, we will require that the operator act promptly to bring the plant within allowable limits.

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We will require the licensee to propose technical specification revisions to conform to this position prior to approval of 100 percent power operation.

We also required that: "The licensee must propose, for staff approval, revisions to the plant technical specifications which will specifically preclude operation at power-to-flow ratios in excess of those for which the plant transient response has been shown to be acceptable".

3.2 Staff Evaluation Subsequent discussions and a letter dated January 30,1981 (P-81032) addresssed the staff's concerns, and the licensee committed to immediately reduce power to lower the power-to-flow ratio to less than 1.17 if a transient results in increasing the ratio.

The licensee also committed to take action for transients involving a power-to-flow ratio less than 1.17.

For power-to-flow ratio exceeding values of Figure 3.1-2 but less than 1.17, an operating time limit of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> has been established to ensure fuel integrity.

If the combination of power-to-flow ratio and percentage of design core thermal power exceeds the curve of Figure 3.1-2, the operator will take prompt action to bring the combination of power-to-flow and percentage of design core thermal power under the curve of Figure 3.1-2.

If this cannot be accomplished in thirty (30) minutes, an orderly shutdown shall be initf ated.

The limitation of allowable operating time to a value of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for all operations with a power-to-flow ratio above the curve of Figure 3.1-2 and below a value of 1.17 provides a conservative limit since this is the allow-able time for a power-to-flow ratio of 1.17 given by Figure 3.1-1.

Based on' the analyses, a 4 (four) hour continuous operating time in this range of power-to-flow ratios would be conservative with reference to possible fuel damage.

However, from an operating viewpoint a 30 minute time limit has been established for operator action, which adds sufficient conservatism.

3.3 Staff Position The licensee has committed to take the corrective steps as outlined in the Evaluation by letter dated March 9,1981.

Therefore, we remove the restriction to operating above 70 percent power as previously stated on page 31 of the Safety Evaluation Report of Amendment No. 22 dated August 19, 1980.

The licensee's commitments have been incorporated into the Technical Specifications dealing with power-to-flow ratios and further action is not required.

, 4.0 RECA-3 Verification 4.1 Introduction The power level of the Fort St. Vrain plant was originally limited to 707, of design power because of limitations in the helium purification system which must be used for depressurization in the event of a loss of forced circulation accident. During reanalyses discrepancies between the values for core region peaking factors and outlet temperature dispersion _used in' the FSAR. safety analyses and the values used in the

. plant technical specifications (which were higher) were identified.

The evaluation presented in the Safety Evaluation Report of Amendment No. 22 addressed the accident reanalyses in support of full power operation.

The licensee has submitted analyses of three accidents which are considered to be the most limiting.

In support of the three bounding accidents, the applicant submitted the results of a review performed for all accidents originally analyzed in the FSAR.

For those accidents affected by either Region Peaking factor or outlet temperature dispersion, a set of enveloping accidents was identified. The staff has reviewed the enveloping logic and the results of the review and found

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acceptable the' conclusion drawn by the applicant that the three accidents identified are_ bounding.

All reanalyses were performed using the RECA3 code.

This code was not used to perform any previous analyses submitted to the NRC (i.e., for the FSAR). While the staff has not reviewed the code for applicability on a generic basis, we have determined the code to be acceptable for the specific-analyses perforced for the Fort St. Vrain Plant.

The staff has determined the acceptability of the applicant's analysis methods by (1) evaluation of key input assumptions to which the output is sensitive, (2) conparison of _the results of applicable plant transient temperature data to temperature predictions for those transients using the RECA3 code, (3) comparison of temperatures predicted by RECA3 toL temperatures predicted by ORECA, and (4) comparison of analysis code-predictions to hand. calculations.

The_0RECA code, which predicts the transient behavior of gas-cooled reactors,- is similar in function to the applicant's RECA3 code.

ORECA was developed by ORNL for the NRC.

The plant data used for code verification were from three reactor trips which occurred from power, and from one event in which all forced circulation was lost for approximately ten minutes.

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' 4.2 RECA-3 Code Verification The RECA3 comparisons to available scram data indicate that predictions of helium temperature in the maximum peaking factor refueling regions are in good agreement with the measured temperature. However, the code underpredicted heliun temperatures in the north-west quadrant of the core by as much as 50*F to 100'F in the 40-70 second time frame. This dis-crepancy may be due to excess bypass flow through fuel region gaps in this quadrant.

Such observations are consistent with region outlet temperature fluctuatic.n phenomena observed during plant operation. The fluctuations were most prominent in this region, and are believed to be due.to the opening and closf ag of axial gaps between fuel blocks.

The discrepancy between the predicted and measured region outlet temper-atures is of concern to the staff. We therefore, required that the applicant perform at least one verification transient subsequent to This corrective action taken to eliminate the core fluctuations.

transient can be a reactor trip from power, and the verification should consist of comparisons of measured to predicted region outlet temperatures. Acceptable predictions of the measured data, including resolution of the previously observed northwest quadrant discrepancies, will be required before full power operation.is allowed. Alternatively, the licensee should identify an acceptable operating power level, based on accident analyses in which this uncertainty has been properly accounted for.

4.3 Evaluation The staff has reviewed the accident reanalysis submitted by the licensee in support of operation of the Fort. St. Vrain plant at 100 percent of design power. Based on our review, we have concluded in the Safety Evaluation Report supporting Amendment No. 22 that the reanalyses provided are acceptable to justify full power operation. However, in our Safety Evaluation in Amendnent No. 22, the staff stipulated that prior to operating at any power level above the present 70 percent restriction, the licensee must perfor:a the.following:

Provide for staff review and approval a minimum of one additional RECA3 code verification analysis of plant

-transient response. The transient response used for verification must be perfonned subsequent to corrective actions taken to eliminate the core fluctuations.

Alternatively, an icceptable power level should be proposed which is Used on accident analyses which account for this prediction uncertainty.

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O, In subsequent discussions with the staff, ACRS, and by letter dated February 5,1981 the licensee has presented a fifth comparison between predicted and measured core region outlet temperatures during cooldown following a scram. The selected transient occurred on July 8,1980, during refueling cycle 2 and subsequent to installation of the region constraint devices (RCD'.s) designed to eliminate core fluctuations.

As in the previous case, the RECA3 core model was initialized with

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" measured" region peaking factors (RPF's). indicated orifice valve positions, estimated core inlet temperature based upon measured circulator helium temperatures and measured power.

Primary loop flow was taken as measured at the circulator and adjusted by -3% to bring all RECA3 steady-state core outlet temperatures into close agreement with measured values. Active core / side reflector flow was calculated utilizing the RECA3 primary loop flow distribution. Following this simulation of core initial conditions, a scram transient from the 35%

power level was initiated. Core inlet temperature was assumed equal to circulator inlet temperature during the transient; and circulator measured flow was corrected as above to obtain core flow throughout the transient.

Af terheat was calculated based upon plant operating history as applied to the FSAR af terheat curve (FSAR Fig. D.1-9).

Selected orifice valves were repositioned during the transient. This

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has been accounted for in making the RECA3 core outlet temperature predictions.

'4esults on a core region-by-region basis have been presented by letter dated February 5,1981. The predicted outlet temperatures show general agreement with the meacured values in all interior core regions. Agree-ment in the outer ring of regions bordering the side reflector is generally good with some improvement in the anomalous behavior previously observed in the northwest corner. However, anomalous behavior is still apparent.

The measured outlet temperature vs. time behavior of region 20 is perhaps the best illustration. Following the reactor scram, the measured outlet temperature may be seen to rise over 50*F during the first 10 minutes of the cooldown. This change cannot be attributed to altered heat generation and may be caused by a flow redistribution. The measured region outlet temperature transient may be the result of changes in the gas flow / temperature environment as modified by the thermal response of the sensing assembly. This behavior is in agreement with plant operating experience in the fluctuation test mode.

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J. To test the effect which the use of " measured" RPF values may have on the RECA verification, the calculated RPF values have been utilized in a second comparison with the July 8,1980 scram. Calculated RPF's are predicted using FSAR diffusion / depletion methods whereas a RPF is " measured" for each region using the common core inlet helium temperature, the sensed region outlet temperature and a region flow inferred from orifice valve position using experimental valve flow characteristics. Measured RPF is the result of a heat balance uiing this data.

Again the predicted outlet temperature shows general agreement in the interior regions, and the retrograde temperature behavior of region 20, in the outer ring, has now been enveloped.

In other regions of the north-west segment of the outer ring, a similar improvement is exhibited although anomalous behavior is still apparent.

4.4 Staff Positicn The presented comparison between predicted and measured core region outlet temperature _s during cooldown following a scram show good agreement for most refueling regions. However, anomalous behavior is still apparent in three regions of the outer ring. This cannot be attributed to altered heat generation and is likely an anomaly caused by a flow redistribution as a result of changes in the gas flow / temperature environment as modified by the thermal response This behavior is in agreement with plant operating of the sensing assembly.

experience in the fluctuation test mode.

In order to fully analyze this anomaly so that conditions for long term steady state operations at full power can be with finality, additional testing and analytical comparisons will be required at reactor powers above 70 percent.

Testing above the 70 percent power level will be subject to limitations impcsed Margins established by by the fluctuation test plan described in RT-500K.RT-500K will ass In addition the use of " calculated" versus " measured" RPFs during testing.to assure that fuel temperatures presented in the FSAR analyses wo be exceeded under possible transient conditions associated with operations above 70 percent power represents an acceptable way of bounding the uncer-tainties represented by the RECA scram comparisons described above for regions in the outer ring.

The. licensee has agreed to submit periodic status reports on this and all In addition, the outstanding issues previously raised during operation.

. licensee has agreed to submit analyses comparing predicted and neasured core region outlet temperatures during cooldown following a scram for review afte each refueling for the next several segments.

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J. 5.0 PCRV Depressurization-Loss of Forced Circulation Accident

5.1 Background

One of the postulated. Design Basis Accidents for the Fort St. Vrain Reactor is a loss of all Forced Circulation Capability (LOFC). In the event of such an accident, and assuming forced circulation core cooling could not be_ restored, it would be necessary to depressurize the reactor to reduce the convective heat load on the upper portion of the Prestressed Concrete Reactor Vessel (PCRV) to assure its integri ty.

Once depressurized, convective heat transfer becomes unimportant, and decay heat can be continuously and safely dissipated to the PCRV Liner Cooling System by radiative and conductive heat transfer.

In the event of permanent loss of forced cooling capability, core temperature would increase slowly and lead to release of fission products to the PCRV due to fuel particle coating failures, however, these fission products would be confined within the PCRV. Depressurization is accomplished by venting the helium coolant inventory-to atmosphere through the Helium Purification System. This system removes fission products from the helium by adsorption in high and low temperature adsorbers so that release of activity during venting is negligible.

However, during core heat up there would be a time dependent increase in the quantity of fission products in the PCRV, and it is necessary to complete depressurization before the capacity of the Helium Purific-ation System is exceeded.

In the FSAR analysis,_ it was established that five hours would be available to restore forced circulation before temperature limits would be exceeded at-the steam generator inlet duct insulation liners, making further attempts to restore forced circulation inadvisable.

Accordingly, five hours after LOFC was chosen as the time at which depressurization would be mandatory.

It was assumed in that analysis that depressurization could be completed in an additional seven hours.

While developing procedures for use with the Alternate Cooling Method,

.PSCo reviewed the capabilities of the as-built Helium Purification System and found some deficiencies which would require modifications and further analysis before it coulo be concluded that depressurization could be safety accomplished following full power operations.

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.i In connection with presently authorized operations at 70 percent power,

.PSCo performed analyses which led to the conclusion that the Helium Puri-fication System would perform satisfactorily if depressurization were initiated at two hours, rather than five, and power was limited to 70 percent. The staff conducted a detailed review and independent analysis to confirm PSCo's conclusion and approved a technical specification change to LCO 4.2.18.

This technical specification stipulates that the reactor shall not be operated at power unless a flow path is available for depressurization to the reactor building exhaust via components of the Helium Purification System and that depressurization must be initiated two hours following onset of LOFC.

Details concerning the PSCo and staff analyses are set forth in the Safety Evaluation issued in support of Amendment 18 to the license.

At the time of Amendment 18, PSCo had' undertaken piping modifications to permit more rapid depressurization (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> as opposed to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with the unmodified system). Tests were done to verify the flow rates achievable with the modified design; procedures were' developed to maximize depressuri-zation flow rates without overheating components; and further analyses were done to establish the capability of the purification system to retain fission products during depressurization, assuming full power operations with an equilibrium core. The objective was to demonstrate that depressurization via the modified system could be safety accomplished from full power equi-librium core conditions. These analyses are reported in a letter from PSCo dated December 22,.1977. Our evaluation of these analyses and the indepen-dent analyses performed by the staff are described below.

By letter dated January 3,1979, PSCo requested a technical specification change which would specify mandatory depressurization times as a function of prior operating history. The purpose of this r.hange is to define conditions under which depressurization would not be mandatory at two hours, as presently specified, thus avoiding unnecessary or premature depressurization which might complicate actions related to restoring forced circulation. Our evaluation of this request is also discussed below.

5.2 Depressurization From Full Power Equilibrium Core Conditions 5.2.1 ~ Performance of Modified System Depressurization of the PCAV under LOFC conditions is accomplished by venting the reactor helium through a train of the Helium Purification System. The flow path includes the High Temperature Filter Adsorber (HFTA), the Helium Purification Cooler, the Helium Purification Dryer, v

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3 Low Temperature gas-to-gas heat exchanger, the Low Temperature Absorber and associated piping and valves to the reactor building exhaust. Previously, depressurization rate was limited by the resistance of the 1-inch vent piping. This piping has been increased to 2-inch in the modified design and includes a 1-inch bypass valve which is utilized to limit flow to the helium purification system during the first 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (+ or - 15 min.) of depressurization.

This prevents overheating of the front-end components of the Helium Purification System during the initial stage of depressurization.

Thereafter, depressurization rate is maximized by opening a 2-inch valve, and depressurization is complete in a total of seven hours.

Prior to initiation of depressurization, the high temperature filter absorber (HTFA) is connected to the Reactor Plant Cooling Water System and a flow of 40 gpm is initiated in the HTFA coils. PSCo's analyses for these conditions, show that the helium exiting the HTFA does not exceed 800'F at the maximum heat load.

At this peak temperature, there will be no loss of integrity, and the HTFA will effectively adsorb todine and particulate activity.

PSCo's review of the helium dryer performance indicates that the effective bed capacity is reduced by a maximum of 40 percent (for C0 )

9 during the initial venting period, when the bed temperature is highei-than normal. Even under these conditions, less than 5 percent of the bed capacity is required to absorb the water and CO2 entering the bed during the depressurization, assuming design concentrations of these impurities (10 ppm H.,1 ppm, CO ).

Therefore, the dryer bed will 20 2

perform satisfactorily under these conditions. Thus the liquid nitrogen cooled Low Temperature Absorber (LTA) downstream will not be affected by the presence of moisture or CO '

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Based on detailed review of PSCo's analyses, we concur that the.ront-end components of the Helium Purification System will maintain integrity and perform satisfactorily under peak load conditions imposed during the initial stage of depressurization. Performance of the LTA is I

addressed below in connection with the discussion of analyses estab-lishing the fission product loading in that component.

In connection with its flow analysis of the Helium Purification train, PSCo developed a nathematical model (TRAIN). Because flow rates, and associated thermal analyses, are dependent on the accuracy of this computer model, the staff required the model to be validated by test. This was done by venting helium through the system at three test pressures with associated measurements of flow rate, temperature and pressure. Test data from two of the tests provided l

w

s. unambiguous data for validation.

The last test, initiated at about 70 psia was ambiguous because the flow meter bypass valve was inadvertently left open.

However, in that test, sufficient pressure data was taken from which to calculate flow rates. The staff has reviewed these data in detail, however, and has concluded that all three sets of <*ata are consistent and provide satisfactory bases for validation ove. the entire flow range for which the model is applied.

Based on these data; PSCo made modifications to the model. We conclude that_ the TRAIN model, as modified, has been verified empirically and that its application to the thermal and flow analyses 'iscussed herein d

is valid.

5.2.2 LTA Heat Loads During Depressurization Based on the re-analyses of pucification system performance, PSCo has conducted a detailed re-analyses of fission product release owing to core heat-up during depressurization.

Basic assumptions were full power operation (105 percent for analysis) with an equilibrium core and core outlet temperature dispersions up to 250*F, consistent with LC0 4.1.7, at the time of an LOFC.

Initial conditions assume an initial failed fuel fraction of 5 percent, consistent with an equilibrium core. Analysis models described in the FSAR were used to predict fission product release from the initial failed fuel fraction and release from additional-fuel particle failures that w'ould occur during core heat-up.

It was assumed that depressurization would be initiated at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following LOFC and that depressurization would be complete and the system r

isolated at 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> (a 7-hour depressurization cycle consistent with performance predicted with the TRAIN model).

The results of these PSCo analyses predict that the peak heat load in the LTA from adsorbed noble gas fission products and daughter products l

from decay would be 28000 BTU /hr. PSCo has evaluated the LTA for this l

peak load and concluded that the absorption bed would retain its l

xenon and krypton burden, without break through, assuming centerline

)

bed temperatures throughout.

Independent analyses by the staff indicate that the LTA bed will retain fission products at a peak load of 35000 STU/hr and possibly somewhat beyond. Thus it is clear that margin exists and break through will not occur. PSCo has also calculated that the integrated fission product heat load on the LTA during depressurization would be 37,600 BTU. Combined with sensible heat load, this gives a total LTA heat duty of 277,000 BTU. This would consume 522 gallons of _the liquid nitrogen used to cool the LTA. Thus the minimum inventory of 650 gallons stipulated in the Technical Specif-cation LC0 4.2.12 is adequate for depressurization.

0

. The staff has conducted an independent analysis of fission product release during depressurization and made a prediction of LTA heat loads using the same initial assumptions used by PSCo in its analysis, i.e.,105 percent power equilibrium core and a 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> depressurization cycle beginning at two hours after LOFC.

Rather than using the analysis methods described in the FSAR, the staff used the fuel failure model suggested in flVREG-0111 " Evaluation of High Temperature Gas Cooled Reactor Fuel' Particle Coating Failure Models and Data" to predict fuel particle coating failures and fission p/oduct release with temperature.

Release rates of noble gases from initially failed particles, at temperatures below the failure threshold in flUREG-0111,

~

were calculated using a release constant of 0.1 per hour. The results of these independent calculations predict a peak LTA heat load of 22,000 BTU per hour and an integrated heat load of 33,600 BTU.

Although the staff analyses give slightly lower heating values than predicted by PSCo, the correspondence is acceptable considering the differences in the analysis models used.

Since the staff model is considered conservative, the results assure that Helium Purification System will perform satisfactorily during depressurization from a full power equilibrium core condition.

5.2.3 Conclusion Based on PSCo's analysis of Helium Purification System performance and the independent review and analysis performed by the staff, we conclude.that the reactor can be operated at full power equilibrium core conditions with reasonable assurance that the Helium Purification System will prevent release of activity during depressurization following a LOFC.

5.3 ^ Mandatory Times for Initiation of Depressurization By letter dated January 3,1979, PSCo requested a change to Technical

-Specification LCO 4.2.18 to provide bases fo' determining, as a function r

of prior operating history, the time at which depressurization would become mandatory. The present LCO 4.2.18 stipulates that depressurization must be initiated at two hours following LOFC regardless of prior history.

The necessity for depressurization is governed by initial core temperature at the time of LOFC, and with temperatures' lower than those at full power the time can be extended.

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. PSCo proposes to incorporate three figures in the basis for LC0 4.2.18 which would give definitive guidance for deciding when l

the reactor must be depressurized. One of these, Figure 4.2.18-1 stipulates the time available as a function of shut-down time if LOFC occurs while the reactor is shut down.

Figure 4.2.18-2 stipulates the time available as a function of average core outlet temperature for operations below 25 percent power.

Figure 4.2.18-3 stipulates the time available as a function of reactor thermal power from 25 to 100 percent of full power.

The last curve requires depressurization at two hours following LOFC over the range of 70 to 100 percent power and adds a maximum of one-half hour relief when extended down to 25 percent power. This curve is consistant with the analyses described in the foregoing section and an independent check of its suitability is not required. Margins available in the Helium Purification System assure that the system will' perform its function following a LOFC over the entire range described.

We have made an independent check of Figure 4.2.18-1 to assure that core heat cp rates are consistent with the delay times given for initiating depressurization from a shutdown ' condition.

In terms of Helium Purification System performance, the delay times are acceptabl e.

However, it is noted that care should be taken not to interpret the delay times as representative of core temperature conditions immediately after shutdown from a substantial power

-(25 percent or higher). The curve applies af ter normal shutdown temperatures have been achieved by forced circulation.

We have made an. independent check of Figure 4.2.18-2, which is to be used for power levels below 25 percent.

This curve is derived on' the basis of average core outlet temperature and is applicable without reference to specific power level, since initial core temperature is derived directly from outlet temperature. We have found that this curve is consistent with core temperature that would be achieved at.various times, following shutdown from temperatures -

derived from outlet temperature, and is compatible with the delay times given to assure proper performance of the Helium Purification System during depressurization.

We conclude the proposed change to LC0 4.2.18 is acceptable and that use of the proposed figures to determine mandatory depressurization times following loss of forced circulation will assure'that the Helium Purification System can perform its function during depressuri-zation without breakthrough and release of fission product activity

-from the absorption beds. This conclusion is conditioned on the interpretation that the delay times depicted in figure 4.2.18-1 are represented on the basis of shutdown time beginning after normal

a. shutdown core temperatures have been achieved through forced circulation cooling.

6.P DEW POINT M0ISTURE PONITORS 61 Introduction The " dew-point" moisture monitors can be used in two modes; to con-tinously measure very low moisture concentrations in the primary coolant system, and to detect rapidly any substantial increase in the moisture level. The instruments use low-temperature nitrogen gas temperatures to chill the surface of a mirror. A continous sample of reactor coolant gas passes over the mirror and causes the mirror to fog, interrupting a reflected light beam if the moisture concentration in the reactor coolant sample increases above a set amount.

The dew point moisture monitors initiate automatic action to mitigate the course of steam generator water-steam ingress events, and are part of the reactor protection system. On detection of high moisture level, a signal is provided to identify and isolate the leaking steam generator from ~its feedwater supply and its outlet steam path. The isolated ~ steam

. generator content is then dumped into a tank to prevent further ingress of moisture into the reactor.

Each primary system cooling loop is provided with three noisture monitor detectors which are identified as the " low level" detectors. Two additional detectors sample-a mixture of gas from both loops and are set at a "high-level" trip point that will cause loop dump and isolation.

The function of the low level monitors is to provide a trip signal which identifies the ' leaking loop. Safety action, which includes dump and isolation of the leaking steam generator identified by the low level

- detectors, is carried'out when one of the two high level -detectors is tripped, also causing a reactor scram.

All detectors receive moisture samples.from a " rake" arrangement of sampling tubes located in the outlet ducts of.each of the helium circulators. The samples from each rake are transported to the moisture detectors.

Af ter passing through the detection chambers, sample gas is returned to the inlet plenum of the circulators. The driving force for'the gas sample depends on the pressure rise across the circulators.

In additionL to this driving force, the sample flow-rate to each detector also depends on line restrictions, including a filter, and the position of a bypass valve that is automatically adjusted to provide an acceptable range of flows across the chilled mirrors.

Flows outside the. acceptable' range are alarmed.

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. Two aspects of the moisture monitors have been reviewed to satisfy the reqirements for 100 percent power operation. The first deals with operability requirements pertaining to Hote (t) to Technical Specifi-cations LC0 4.4.1.

This limiting condition of operation establishes a nininum sample flow as a function of reactor power. This requirement guards against unacceptable degradation of sample flow rate, attributed primarily to potential filter clogging. The filters protect the mirror surfaces against particulates such as graphite dust, and are changed whenever necessary to ensure arequate sample flows. The basis for specifying minimum sample -flow requirements for the dew point moisture detectors is to assure that they will respond prior to the backup safety action from the reactor high primary coolant pressure trip following a design basis steam leak - thus providing maximum assurance of isolation and dump of the correct (leaking) loop.

The second aspect concerns test data on response times for loop selection and trip.

In this regard, we reviewed moisture injection test data obtained at 5 and 25 percent power for consistency with the response times required at 100 percent power for the design basis moisture ingress accident. The licensee was committed to perform a moisture injection test at 100 percent power to confirm response characteristics.

However, this requirement was modified to require that the licensee perform a noisture injection at approximately 70 percent power to provide additional confirmatory information. A test at that power will provide essentially the same confirmatory response time information as a test t higher power.

6.2 Sample Flow Limits for the 70 to 100 Percent Power Range By letter dated May 31, 1977, the licensee proposed a number of changes to LCO 4.4.*e to extend the specification to cover power levels up to 100 percent of rated power. However, action was not taken on that request, pending results from moisture injection tests at about 70 percent power, and that request was withdrawn.

In a subsequent request dated November 16, 1977, LC0 4.4.1 Note (t) extended minimum DPMt1 sanple flow rates to 70 percent power, with a minimum specified sample flow rate of 50 scc /sec over the range of 35 to 70 percent power.

It does not extend to full power. Over the power range from 70 to 100 percent the nominal flow rate would be 62.5 sec/sec. To extend the specification to full power, the licensee has proposed, durf 1 several discussions, that the minimum sample flow rate over the range '

70 percent to full be specified at 50 scc /sec.

This would allow operatioi atitude for reduction in the nominal sample flow rate owing to cloggint le filters without impairing requirements 1

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=, related to trip functions (loop identification, scram ana steam generator dump) within response time requirements.

6.3 Staff Position We have again reviewed test data from moisture injection tests done at 5 percent and 25 percent power, together with response characteristics for the moisture monitor system reported in GA-A13823 and have concluded that DPMM -response characteristice extrapolated to the 70 to 100 percent power range will be consistent witn requirements for response times and loop identification if a minimum sample flow rate of 50 scc /sec is specified for that range. Therefore, we conclude that Note (t) to LCO 4.4.1 can be modified accordirgly.

Further modification of LC0 4.4.1 to incorporate flow limits similar to those in the withdrawn request, (with alarm ranges governed by the interlock sequence switch) will not be made until the 70 percent moisture injection test results are available and the previous.'equest is re-submitted. At that time we will determine whether the proposed minimum specified sample flow rate of 40 scc /sec over the 30 to 100 percent power range is acceptable.

In the meantime, adherence to a minimum flow of 50 sec/sec will assure acceptable response characteristics.

7.0 Plant Testing Above 70 Percent Rated Power 7.1 Introduction Public Service Company of Colorado plans to continue their rise to power

. testing.from 70 to 100 percent power in ' order to:

(1) obtain systens performance _ data over -that power range, (2) complete the previously authorized start up tests, and (3) obtain further test data on reactor performance with the installed region constraint devices.

After the planned tests are completed, PSCo will operate the reactor at power levels not in excess of 70 percent until the staff has reviewed the

- test data to ensure that the operating limits based on the test data are consistent with safety related requirements for long term, steady-state operation at power levels greater than 70 percent rated power.

Testing of the cycle 2 core without the region constraint devices (RCD) indicated that the fluctuation behavior of the core was similar to that observed in cycle 1 although with some alteration in' detailed-characteristics. Region constraint ' devices were then ' installed.to stabilize the top plane of the core by ' mechanically tying refueling regions together. Testing done in November and December, 1980, under conditions that would have triggered fluctuations.in' an unrestrained

. 3 core, confirmed that fluctuations would not occur with RCDs in place under the most severe test conditions imposed. However, with a relatively high pressure drop across the core, temperature redistribution occurred which resulted in an increase in the outlet temperature of central refueling regions and a decrease in peripheral regions. This behavior has been attributed to a small redistribution of core flow due to a hydraulic clamping force at the core boundary which decreased inter-region gap clearances in the inner regions of the core-and correspondingly increased gap clearances at the boundary.

The net result is a slight reduction in cooling flow to central regions, an increase in by-pass flow at the boundary, and a corresponding increase in measurement error for region exit temperatures.

7.2 Staff Evaluation Analyses by the licensee have shown that when the temperature redistribution occurs, the overall temperature response is consistent with the mechanism which has been postulated.

The redistribution behavior was reproducible in both the November and December tests and once it occurs, is apparently stable at higher power levels. There is no tendency for fluctuation after redistribution occurs, and it appears that stabilization by the RCDs at the top plane of the core has corrected any tendency for cyclic changes in interregion gap dimensions that has been associated with fluctuation in the unstab-ilized configuration without RCDs.

For testing ab've 70 percent power, PSCo has revised their testing procedure (RT 500K) to take advantage of experience gained in previous testing.. In addition,- the revised test plan incorporates staff comments for the establishment of margins on outlet region temperatures, conditions before and after flow redistributions, should they occur, and te:nperature uncertainties as needed to assure that Technical Specification limits are met at all times.

The testing strategy has also been altered. The November and December tests were performed.on system operating lines which gave artificially-high pressure drops at any given power level so that previously.

estab.lished fluctuation thresholds would be exceeded. By using the high pressure drops, the orifice configuration was artificially skewed into an abnormal pattern resulting in a configuration not likely to be encountered during normal operation. For the testing above 70 percent power, the orifice configurations for normal operation will be used.

The tests wi.ll be initiated at approximately 40 percent power and will proceed in approximately 3 percent power increments. If a temperature -

redistribution.occ'trs, system test margins will be reestablished according

. '. to prescribed procedures in order to assure that sufficient margin is maintained to meet technical specification limits on region outlet mismatch temperature as power is increased.

These procedures, as revised following discussions with the staff and their consultants, conservatively take into account uncertainties related to measurement errors, including gap flow interference with ' outlet thermocouples and outlet temperature extrapolation to full power. _ From our review of these procedures, we are satisfied that the tests can be done without exceeding LC0 limits. While fluctuation behavior is not expected, based on the results of previous tests and associated analyses, provision has been made in the test plan for acquisition of data should fluctuations be encountered.

The amount of time in a fluctuating mode will be limited to approximately 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per run.

7.3 Staff Position We have reviewed the proposed test plan RT-500K and found it to represent the actions that PSCO plans to take in performing rise-to-power and fluctuation testing. The plan stipulates that the testing will be performed on the original system line that was used prior to encountering the 1977 fluctuation phenomenon. The limits imposed are such that the test results do not exceed the Tcchnical Specification limits. The limits imposed are consistent with those previously established to assure that design limits for equipment are not exceeded and a. fundamental limitation throughout

.is adherence to Technical Specification limits.

Basic revisions to.the test plan RT-500K were made by PSCo following discussions'with the staff:

(1) The scale up of nuclear' channel deviations has been changed from 20 percent-to ~15 percent for the range above' 70 percent power to more accurately reflect.the maximum expected deviation in that range.

(2) The allowable' reactivity anomaly has not been addressed since the flux control' system will' operate automatically to immediately

ompen ate for any reactivity changes which could be experienced.

Calculations have shown that complete closure of all vertical gaps in the core, if it could occur, would result in a reactivity insertion of about 3 cents (0.00015ap).

(3).Both P5Co 'and GAC stated that the re:alculation of temperature mismatch would be performed by a computer program provided for that

-particular purpose.

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. (4) The formula used to calculate the temperature mismatch has been revised to correct the mismatch for a projected power increase.

(5) To account for the possibility of outlet thermocouple errors, due to gap flow penetration of thermocouple sleeves in the outer regions, physics calculations will be performed to compare

" expected" and " measured" temperatures. These calculations will start at 40 percent power and will be repeated at each 10 percent power increase.

Those regions that exhibit a difference of -70*F or more will subsequently have outlet temperatures adjusted by comparison with regions which are not subject to such thermocouple error. The remaining regions will be governed by the 4

margins indicated by Figure 1 of RT-500 until such time that a temperature redistribution occurs. This modification accomodates the staff concern that operations based entirely on margins in Figure 1 might overlook the possibility of thermocouple errors in some outer regions far which the initial margins would not be sufficient.

(6) The procedure for adjusting margins af ter a temperature redistri-bution has been modified-to provide verification by physics calcu-lations. Also, an additional margin of 25*, for a planned power increase of 5 percent, or 50', for an increase of 10 percent, has been added to the formula used to derive ' margins after a redistri-

~

bution. _ At each 5 or 10 percent increase in power the physics calculations will be repeated to provide further ' adjustments as needed. When a temperature redistribution occurs, the procedure will be used 'to derive interim. margins calculated on the basis of measured parameters. These interim' margins will be used for operations at the power level at which the. redistribution occurred and.further rise-to-power will not be performed until margins based on physics. calculations for each region governed by the procedure have been established.

It should be noted.that Figure C on page.27 of RT-500K presents a more

-restrictive calculation _of margins that.those -in the test procedures.

Figure C is part of the plant Standard Operating Procedures and is used as'a guideline in the test procedure and not as a restriction.

All of the staff comments have.been incorporated into the revised test plan RT-500K. We-conclude that with adherence to the limits and methods described in the plan, the testing program above 70 percent power' can be performed as proposed without exceeding design and. Technical Speci-fication limits..

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d 8.0 Staff Conclusions Based on our review of the documentation referenced in this report, an evaluation of plant operations thusfar, evaluations of the plant through site visits by NRC technical specialists, and favorable reports by the NRC Office of Inspection and Enforcement, we conclude that the Technical Specifications can be revised as requested. This safety report describes the basis for this conclusion, and notes the conditions which apply.

Since full power operation has previously been approved,

.this amendment presents the staff's continued approval of certain items for full power operation.

The staff has determined that the issues addressed in this amendment will not result in any significant safety or environmental impact not previously evaluated.

Since the Fort St. Vrain reactor is the first and only plant of this size and type, and since a substantial base of experience comparable to that for light water reactors does not exist, the performance of the Fort St. Vrain reactor continues to be closely monitored by NRC staff.

As.related in Amendments No.18-and No. 22, three items formed a formal hold to operation of the reactor _ above 70% power:

(1) Depressurization, (2) Moisture Monitors, and (3) Accident Reanalysis.

The third item has been reviewed and our comments and conclusions are presented in Section 9 of-the Safety Evaluation Report in Amendment No. 22. _ This amendment presents our evaluation of the licensee's response to the.

comments relating to accident reanalyses. Also presented are our evaluations dealing with the other items that constituted a hold in the rise-to-power operations to 70 percent poser.

This Safety Evaluation Report presents the staff's approvit for continued rise-in-power -testing and fluctuation testing to the designed limit.of 842

~ Wt.. Testing above the 70 percent power level will be subject to limi-tations imposed by the fluctuation test plan described in RT-500K..This testing should be ' completed prior to July 1,1981. After the test program is completed, the plant may not be operated above 70 percent power until the staff has reviewed and approved the test results and assured that satisfactory. bases exist for prolonged steady state operations above 70 percent.of full rated power.

The staff has reviewed the March 4,1981 revisions to the test plan RT-500K

- and the test plan proce,dures. The ' test plan procedure RT-500K, as revised on 11 arch 4,1981, states the actions that are' planned during the fluctuation testing up to full rated power of 842 MWt.

The test plan conforms to requirements made by the NRC staff and is acceptable.

D 7

. The staff will require status reports on this fluctuation testing.

In addition, analyses comparing predicted and measured core region outlet temperatures during cooldown following a scram shall be presented for review af ter each refueling for the next several segment reloads.

Dated:

,.gg, 6 1991

APPENDIX A i

CHRONOLOGY OF FORT ST. VRAIN LICENSING ACTIONS PERTAINING TO PLANT OPERATION, SAFETY EVALUATIONS AND LICENSE AMENDMENTS DATE TITLE September 17, 1968 Commission issued a construction permit for the Fort St. Vrain Nuclear Generating Station.

November 4,1969 Public Service Company of Colorado submitted the FSAR as amendment 14 to its application for a construction permit and license.

January 20, 1972 Safety Evaluation by the Division of Reactor Licensing, U. S. Atomic Energy Commission in the matter of Public Service Company of Colorado - Fort St. Vrain Nuclear Generating Staticn, Docket No. 50-267. This document pertained to the review of the Final Safety Analysis Report prior to issuance of an operating license.

June 12, 1973' Supplement No.1, Safety Evaluation by the Directorate of Licensing,'U. S. Atomic Energy Commission in the

- matter of Public Service Company of Colorado - Fort St. Vrain Nuclear Generating Station, Docket No. 50-267.

This docunent pertained to postulated high energy pipe ruptures outside containrent.

- Decembe r '21, 1973

_ License No. DPR-34 issued for the operation of the Fort St. Vrain Nuclear Generating Station.

May 17,l1974 Safety Evaluation by the Directorate of Licensing Supporting Amendment No. 1 to License No.'DPR-34.

Changes the Technical Specifications by:

(1) making exceptions to requirements for installation of secondary.

-closures during certain initial low power physics testing, (2) revising specifications for monitoring during certain radioactive ' effluent releases, -(3) _ revising specification for tendon load cell and PCRV concretc i

crack surveillance, (4) revising _ certain specifications l

for _ checks,_ calibrations, and testing of loop shutdown system, and (5) redefining certain administrative responsibilities and authorities of the offsite Nuclear Facility Safety. Committee.

June 27, 1974 Safety -Evaluation by the Directorate ofaticensing Supporting Amendment No. 2 to License No. OPR-34 Changed the Technical Specifications to revise the organization of personnel for Fort St Vrain Nuclear

-Generating Station.

' July 12,1974 Safety Evaluation by the Directorate of Licensing Supporting Amendment No. 3 to License No. DPR-34.

Changed the Technical Specifications to allow low power reactor operation with a helium environment in the' reactor _during Phase I of the power ascension program.

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. 1 Date T_i tl e November 11, 1974 Safety Evaluation by the Directorate of Licensing Supporting Amendment No. 4 to License No. DPR-34 Changed the Technical Specifications to permit revision of (1) radial power peaking factors under certain operating conditions and (2) the number of core regions allowed the maximum deviation in outlet temperature from the average core outlet temperature.

1 December -19, 1974 Safety Evaluation by the Directorate of Licensing Supporting Amendment No. 5 to License No. DPR-34.

Changed the Technical Specifications to permit revised staffing requirerents for plant operating shifts.

-January 23, 1975 Safety. Evaluation by the Division of Reactor Licensing, Supporting Amendment No. 6 to License No. DPR-34.

Changed the Technical Specifications to permit a change in calibration frequency for one adjustment of the wide range power instrumentation and added a calibration requirenent for the linear range power instrumentation.

April 17,~1975' Safety Evaluation by.the Office of Nuclear Reactor Regulation Supporting Arendment No. 7 to License No.

DPR-34 Changed the Technical Specifications to permit bypass of the two-loop trouble scram when the reactor node switch is in the " fuel loading" position.

December.1, 1975 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. S to License No.

DPR-34 Permitted the-possession. d use of additional radioactive sources for the purpose of calibration and

. instrument checks.

Decembdr 29, L1975 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 9 to License No.

0 DPR-34 Changed the Technical Specification to permit a reduction in the helium circulator high-speed trip when operating on water-driven Pelton turbines.

Janua ry -27,1976 Safety Evaluation by the Office of Nuclear Reactor

Regulation Supporting Amendment No.10 to License No.

i-DPR-34.

Changed theLTechnical Specifications to permit

- a change in the procedures to be followed in'the event of trouble with the hydraulic power _ system.

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[e is' 1 Date Title April 15,1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.11 to License No.

DPR-34 Changed the wording in the Technical Specifi-cations to eliminate an inconsistency in the plant protection system labeling and the Final Safety Analysis Report.

April 26, 1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendnent No.12 to License No.

DPR-34.

Changed the Technical Specifications to add surveillance requirements for helium circulators and helium circulator Pelton wheels.

June.18, 1976 Safety Evaluation by the Office ti Nuclear Reactor Regulation Supporting Amendment No.13 to License No.

DPR-34. Changed the Technical Specifications to:

(1) add requirements for operation of analytical system moisture monitors between reactor shutdown and 5 percent power; also calibration frequency for.these monitors is stated; (2). revise allowable. primary system impurity levels and method'of specifying moisture impurity from parts per million to dew point temperature; (3) add a definition of'ooerable dew point moisture monitors; (4) add functional checks and tests for dew peint moisture surveillance

-monitors; (5) revise the core reactivity status and limiting conditions for operation; (6) isolate'the

. helium storage systen from the helium circulator buffer helium system when the reactor-is in operation; (7) allow bypass 'of plant protective system moisture monitors for testing during the startup testing program; and~ (8) add reporting requirements.

June 18. 1976

. Safety Evaluation by the Office of Nuclear Reactor Regulation supporting amendment.no.14 to licensee no. DPR-34. ' Revised the Technical Specifications to add requirements for: (1) backup pumping capability

-to the fire water system; (2) surveillance for the.added pumps; and (3) an additional class lE power source.for the plant protective system.

June 24, 1976 Sa fety Evaluation by. the 0ffice of Nuclear Reactor Regulation Supporting Amendment-No.15 to License No.

DPR-34. - Changed Technical Specifications to add requirenents for operability and surveillance of shock suppressors.

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Date Title November 17, 1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.16 to License No. DPR-34.

Revised the section of the Technical Specifications relating to administrative controls.

December. 8,1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.17 to License No.

DPR-34.

Temporarily revised the provisions in the Technical Specifications relating to operation of the bearing water makeup pumps in the primary coolant system.

October 28, 1977 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.18 to License No.

DPR-34.

Permitted Stage 2 operation up to 70 percent of rated thermal power.

February 23, 1979 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 19 to License No. DPR-34.

Incorporates the Fort St. Vrain Amended Security Plan as part of the license.

April 20,1979' Safety Evaivation by the Office of Nuclear Reactor Regulation supporting amendment no. 20 to license no. DPR-34.

Revised the Technical Specifications to:

(1) install eight test fuel elements into the reactor core at the first refueling, and (2) install _ PGX graphite surveillance -specimens into five bottom transition reflector. elements of the reactor core.

June 6,1979 Safety Evaluation by th Office' of Nuclear Reactor Regulation supporting amendment no. 21 to license no. DPR-34.

Revised the Tech.1ical Specifications to:

(1) modify the firei protection system for the three room complex, the Auxiliary Electric Room, the 480 Volt Switchgear Room and the congested cable areas; this constitutes Stage III fire

. protection implementation; (2) convert the Interim Alternate Cooling method to the final Alternate Cooling Method; (3)'

test the. reactor building louver. system on a quarterly basis; (4) eliminate the manual isolation of the high pressure helium supply from the _ helium circulator buffer supply header; and (5) add two firewater booster pumps to the firewater system to provide adequate capacity to operate a circulator water turbine and supply emergency cooling water for safe shutdown cooling.

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- Date Title

~ August 19, 1980 Safety Evaluation by the Office of Nuclear Reactor Regulation supporting amendment no. 22 to license no. DPR-34.

Revised the Technical Specifications to (1) change the

-amount of diesel fuel in each diesel generator set day tank to -325 gallons; (2) update company reorganization based on NRC requirements;-(3) change the number of hours

'that the ACM diesel generator can operate with 10,000 gallons of fuel to 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br />; (4) alter the Fire Protection Technical Specifications to_ follow the requirements of STS on Fire Protection; (5)- change the frequency and method of

~ Reactor Protective System Surveillance to satisfy the requirement of IEEF-279-1971; (6) update the listing of all snubbers; (7) change the fissile particle thorium to uranium ratio to reflect "as manufactured" specifications and (8) change the-values for core region peaking factors and outlet temperature dispersions to reflect existing values ir. conjunction with accident. reanalyses in support of full power operation.

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-s.--