IR 05000346/1979004
| ML19261D642 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Crane  |
| Issue date: | 03/27/1979 |
| From: | Chow E, Creswell J, Streeter J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19261D640 | List: |
| References | |
| TASK-TF, TASK-TMR 50-346-79-04, 50-346-79-4, NUDOCS 7906200486 | |
| Download: ML19261D642 (15) | |
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U.S. NUCLEAR RL ULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
Report No.:
50-346/79-04 Docket No.:
50-346 License No.:
NPF-3 Licensee:
Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, Ohio 43652 Facility Name: Davis-Besse Nuclear Power Plant, Unit 1 Inspection At:
Davis-Besse Site, Oak Harbor, Ohio Inspection Conducted:
February 5-9, 1979 LC
/, L. su ~ -. < A-
IO 7 Inspectors:
J. S. Creswell Tde (""'*)
3/2.3 b7 E. T. Chow (Date)
e
__
P
. v. T. -
J. F. Streeter (February 9, 1979)
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.N "~U J-s y., /, e Approved By: J.' F. Streeter, Chief
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t Nuclear Support Section 1 (Date)
Inspection Summary Inspection on February 5-9, 1979 (Report No. 50-346/79-04)
Areas Inspected:
Routine, unannounced inspection of core thermal margin determination; power coefficient of reactivity determination; core thermal power evaluation; rod worth measurements; shutdown margin determination; unresolved items; audit of the Inapector (J. S. Creswell). The inspection involved 79 inspector-hours onsite by three NRC inspectors.
Results: Of the six areas inspected, no Items of Noncompliance or Deviations were identified in 5 areas. One Item of Noncompliance was identified in one area (Infraction - failure to follow procedures, Paragraph 4).
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DETAILS, 1.
Persons Contacted
- T.
Murray, Station Superintendent G. Novak, Superintendent, Power Engineering C. Domeck, Project Engineer
- B. Beyer, Assistant Superintendent
- S.
Quennos, Technical Engineer
- D.
Lee, Test Program Manager F. Miller, Power Engineering Engineer The inspectors also interviewed other licensee employees, including members of the technical and operations staf f.
- Denotes those attending the exit interview.
2.
Licensee Actions on Previous Inspection Findings (0 pen) Unresolved item (Inspection Report 50-346/78-06, Paragraph 4):
Power oscillations. The licensee is presently collecting data to determine the magnitude of power oscillations.
(Closed) Unresolved Item (Inspection Report 50-346/78-06, Paragraph 4):
BTU limit calibration.
Flow calibrations were performed on June 2 and October 16, 1978.
(Closed) Unresolved Item (Inspection Report 50-346/78-06, Paragraph 5):
Incore detector calibration.
Calibration Procedure IC 2001.06 was completed on March 23, 1978.
(Closed) Unresolved item (Inspection Report 50-346/78-06, Paragraph 5):
Background correction factors.
Background correction factors were implemented on July 29, 1978 (except for String 8 which is inoperable).
(Closed) Unresolved Item (Inspection Report 50-346/78-06, Paragraph 5):
Aluminum oxide insulated detectors.
Software has been modified to allow treatment of this type of detector.
(Closed) Unresolved Item (Inspection Report 50-346/78-06, Paragraph 6):
The inspector notified the licensee that Region III had decided the item was not reportable.
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(Closed) Noncompliance (Inspection Report 50-346/78-06):
Failure to obtain proper review and approval of changes to Procedure TP 800.29.
Rod drop test procedure changes were reviewed and approved on June 17, 1978.
(Closed) Inspector Followup Item (Inspection Feport 50-346/78-06, Paragraph 9)-
RCS pressure instrumentation.
The licensee stated that the problem was traced to a faulty channel in the reactimeter which was repaired.
(Closed) Inspector Followup Item (Inspection Report 50-346/78-06, Paragraph 9): Monitoring of rod control system.
No further monitoring was reported.
(0 pen) Inspector Followup Item (Inspection Report 50-346/78-17):
Review of TP 401.01.
Power Engineering completed its review on July 31, 1978, but the comments have.io t been resolved.
(Closed) Inspector Followup Iter (Inspection Report 50-346/76-17, Paragraph 6):
Determination of conservative FA P Values.
Document-ation from the vendor indicates that the central assembly H-8 was the peak power assembly. Although uncertainties in the evaluation do not allow a determination of the precise value for FaF, the evsluation indicates FoH was probably very close to the Technical Specification limit.
(Open) Unresolved Item (Inspection Report 50-346/78-17, Paragraph 15):
Rod drop test evaluation. The inspectors recuested the following information:
a.
Compare the results of PDQ and online computer results for dropped rod test conditions.
(This is effectively a test case to verify the capability of the online computer to accurately or conservatively measure core parameters.)
b.
Provide constants for the curve fitting algorithm used in the peak to segment power calculation.
c.
Compare the PDQ and online computer results for the following values:
(1) Axial Local Peak 228:6
- 29 (2) Radial Local Peak (3) Radial Peak Uncertainty-3-
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(4) Axial Peak Uncertainty (5) Peak to Average Segment Power (Closed) Noncompliance (Inspection Report 50-346/78-17):
Failure to follow procedures. A Required Reading List concerning the event was sent out September 12, 1978.
The licensee had included procedure changes to address the commitments related to determining worst case thermal conditions.
(0 pen) Unresolved Item (Inspection Report 50-346/78-27, Para-graph 2): High pressure injection performance. The inspectors met with Power Engineering personnel on February 7, 1979, to discuss inspector comments related to the High Pressure Injection (HPI) review performed by Power Engineering.
The inspector noted that a flow versus pressure comparison of the September 1977 event and the small break analysis had not been made. The licensee stated that preoperational test results and consideration of instrument errors verified HPI system operability. The licensee also stated that NRR had reviewed the September 24, 1977, event and by its review had approved the HPI operability issue.
The inspectors requested the following:
a.
What HPI timing / flow assumptions were made in the small break analysis?
b.
What testing has been or will be done to demonstrate the HPI pumps will be delivering on their head curves within 30 seconds of the initiating signal?
(Most limiting case is simultaneous LOCA and Loss of Offsite Power, time to sequence HPI pumps on the emergency diesel generators, and time for HPI pumps to come up to flow on their head curves.)
c.
If stop check valves in 3 of the 4 HPI legs " stick" following closure of the valve (s) because of maintenance or testing, could the flow delay caused by the " sticking" result in HPI pumps not delivering flow in accordance with accident analysis assumptions?
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(0 pen) Unresolved Item (Inspection Report 50-346/78-30-02):
Review of natural circulation test results.
The inspector's question regarding a lower natural circulation flow at the 60 inch steam generator level than at the 35 inch level was discussed with test leader. The licensee maintained that instrument errors on the order of 3% are responsible for this anomaly.
Test data indicates that natural circulation flow is relatively insensitive to auxiliary feedwater level in the steam generators over the startup level range measured (95 inches to 35 inches). The variation in natural circulation flow for the range of levels was approximately 0.1% of full reactor coolant system flow. The explanation of the relatively constant flow was that the auxiliary feedwater is sprayed directly on the tubes from a location high in the steam generator such that the majority of the primary to secondary heat transfer takes place in the area where the auxiliary feedwater impinges on the tubes and then flows down the tubes.
The natural circulation flow values determined are for steady state conditions.
For example, the 35 inch level would produce an equilibrium RCS average temperature of 566 F.
For a steam generator level of 120 inches (auto essential level setpoint), an equilibrium RCS average temperature of 552 F is predicted.
For both cases (35 inch and 120 inch levels), a T cold greater than 530 F is predicted.
Since this data was used to support administrative control of auxiliary feedwater level (with no small break), the inspector had the following comments about the data:
a.
Values of T hot for levels below 35 inches are not supported by measured values, b.
The data was taken at equilibrium conditions. Transient effects on cooldown rate were not defined in the data.
c.
Effects of overshoot or undershoot during manual control at the 35 inches level are not clear.
The inspector will review test data for testing performed subsequent to the natural circulation test when administrative controls were employed.
This review will be to determine the effect of manual auxiliary feedwater control at the 35 inch level.
(0 pen) Unresolved Item (Inspection Report 50-346/78-30-03): Technical specification change.
The change had been approved by the Safety Review Board and had been sent out for QA review.
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(0 pen) Unresolved Item (Inspection Report 50-346/78-30-04): Core lift analysis. During discussions with Power Engineering Personnel on February 7, 1978, the inspector was informed that the subject questions had been forwarded to B&W for resolution.
The licensee stated that for Cycle 1A (prior to BPRA removal) there was a net upward force on some of the fuel assemblies at full RCS flow (112%).
The licensee stated that for Cycle IB there was less upward force on the assemblies due to flow redistribution.
(0 pen) Unresolved Item (Inspection Report 50-346/78-30-07): Operation at power on auxiliary feedwater system. The licensee stated that the hypothesized event was covered by the loss of feedwater transient.
The inspector stated that one of the purposes of a power ascension test program was to verify the analysis in the FSAR. He further stated that because of events and testing, it is not clear whether or not the analysis performed bounded the hypothesized event.
The inspector noted that the loss of feedwater event assumed a high pressure reactor trip whereas the hypothesized event would apparently not include that automatic action. This issue will be reviewed further.
3.
Core Thermal Margin Determination The inspector reviewed records, items associated with the determin-ation of pertinent core parameters, and discussed these items with licensee personnel. The inspector reviewed the pertinent procedures and data for the month of January, 1979.
No items of noncompliance were identified related to surveillance of axial imbalance, hot channel factors, quadrant power tilts or incore detector channel check.
The inspector reviewed with the Technical Engineer the processing of surveillance tests and the personnel associated with the testing.
Surveillance tests are scheduled using computer printouts.
Surveill-ance Test Engineer is responsible for the distribution of these schedules and the overall coordination of the surveillance test program. He is responsible for assigning Test Engineers to perform the test and he reports to the Technical Engineer.
The Test Engineer secures a copy of the test procedure from the control room control copy file.
The Shift Foreman assigns operations personnel to assist in the testing if required. The Test Engineer reviews the data against the applicable acceptance criteria and notifies the Shift Foreman of the results. The test document is then forwarded to the Designated Reviewer who performs a second review. The test document is then sent to be filed.
The inspectors noted that in 2286
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some instances the Surveillance Test Engineer, Test Coordinator, Test Engineer, and Designated Reviewer were the same individual.
However, the licensee stated this is not routinely the case.
No items of noncompliance were identified.
The inspector noted some improvements in the knowledgeability of the technical staff with regard to the functioning of the Ir. core Detector System. However, concerns regarding the performance of " Physics" testing and technical evaluationgjas documented in previous inspec-tion reports the August 16, 197&- management meeting remain.
Specifically, those concerns related to not following procedures and to staf fing weaknesses in the core physics area.
However, no items of noncompliance were identified in the technical engineering staffing.
Since the Computer Engineer was not at the plant during the inspection, items concerning the programming of the process computer will be addressed in a future inspection.
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The inspector requested information regarding NRR's review of the process computer software. The licensee could not furnish this information but has requested the information from the vendor.
This is an Unresolved Item (50-346/79-04-02).
During the inspector's review of incore detector performance, the licensee stated that Detector String 46 was not replaced during the plan. autage for burnable poison removal.
Apparently there was blocnoge in the guide tube for this string.
Since the string was inoperable, it appeared analytical data was being used to predict exposure, isotopics, etc., for the related assemblies. The inspector stated that the use of such analytical data rather than measured data would result in additional uncertainity in physics calculations for future reloads. This concern is based on the consideration that the fuel now on the periphery would be shuffled into higher power locations in the interior of the core.
The inspector determined that there was no Technical Specification requiring this detector string to be installed.
4.
Review of Power Coefficient of Reactivity Determination On September 14, 1978, the licensee performed Test Procedures TP 800.05, " Reactivity Coefficients at Power" and TP 800.20, " Rod Worth Measurements". The testing was performed at approximately 90% power. The purpose of TP 800.05 was to determine the temperature coefficient of reactivity and the power doppler coefficient. The purpose of TP 800.20 was to measure differential control rod worth.
-1/ Inspection Report 50-346/78-23.
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The inspectors reviewed the control copy of the test procedure, charts, and records and interviewed the Test Leader.
a.
Test Conduct (1) Nuclear instrument and Tave strip charts indicate TP 800.05 and TP 800.20 were conducted as follows:
Prior to the performance of the first four differential control rod worth measurements per TP 800.20, the reactor power was approximately 92%. The duration of time between the first four measurements was approximately 5 minutes.
Af ter these measurements were coupleted Tave war decreased approximately 3 F to 578 F and power decreased approximately 3% to 89%. Measurements 5 and 6 were taken under these conditions. The Tave was then increased approximately 6 F to 586 F and power increased approximately 6% to 95%.
Measurements 7 and 8 were taken under these conditions.
The Tave was then lowered approximately 4 F to 582 F and power decreased approximately 5% to 90% and measurements 9 and 10 taken approximately 5 minutes apart. Power was then decreased approximately 5% to 85% gith an agcompany-ing decrease in Tave of approximately 1 F to 581 F.
Measurements 11 and 12 were then taken approximately 5 minutes apart. Power was then increased approximately 5%
to 90% power. No further rod measurements are indicated.
Examination of TP 800.05 revealed the f ollowing inf urmation:
Regarding the purpose of the test
"the temperature coef ficient will be measured by lowering the Reactor Coolant System (RCS) coolant temperature while holding power constant".
Prerequisite 6.6 states "The reactor is operating at the power level (+ 0.5%) and control rod configuration specified
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by the Power Escalation Controlling Procedure, TP 800.00,
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for 10 minutes prior to measurements.
...
Prerequisite 6.7 states "Tave is constant within 1 F for
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10 minutes prior to the measurements
....
Step 7.1.7 etates "Re-establish (after the Tave decrease)
the stable power, temperature, and pressure conditions specified in Section 6.6, 6.7, and 6.8 of this procedure".
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The strip charts indicate that:
Prerequisite 6.7 was not satisfied before Tave was increased in Step 7.1.8.
Prerequisite 6.6 was not satisfied before Power was increased in Step 7.1.8.
Technical Specification 6.8.1.c requires that procedures be followed. Failure to follow TP 800.05 on September 14, 1978, by not satisfying Prerequisites 6.6 and 6.7 constitutes an example of an Item of Noncompliance (50-346/79-04-01) of the Infraction level against Technical Specification 6.8.1.c.
(2) The performance of TP 800.05 requires that data be obtained per TP 800.20, " Rod Worth Measurements." Review of rod worth data revealed that the licensee identified the following differential rod worth measurements as unacceptable:
First-rod motions not identical Second-rod motions not identical Fourth-rod motions not identical Sixth-rod motions not identical Ninth-rod motions not identical Tenth-equilibrium not good, change just prior to motion The third, fif th, seventh, eighth, and eleventh measurements were found acceptable by the licensee.
The Test Leader stated that the bad measurements were due to an inexperienced reactor operator manipulating the rods. The inspector noted that two measurements are called for when differential rod worths are measured as delineated by TP 800.20.1, Enclosure 2.
The intent of this procedure was to average the results of two measurements to produce a more accurate value.
Due to the large number of unacceptable measurements, average values were not always obtained. However, the Test Leader pointed out that the procedure allowed him to not consider obviously bad measurements.
The Test Leader also stated that the prerequisites as delineated by the procedures were not always followed since he sometimes made a judgment as to when the intent of the procedure was satisfied.
(The Test Leader was a coauthor of the procedure.) The Plant Superintendent was informed by the inspectors that the Test Leader's philosophy regarding following procedures was unacceptable and that procedural controls for changes to procedures must be used when a procedure is not going to be followed.
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b.
Licensee Reviews of Test Results The following chronology of events was constructed from information contained in the control copies of TP 800.05 and Power Engineering comments on the test:
October 2, 1978 Request for Review and Comments sent to Power Engineering (comments requested by October 16, 1978)
October 3, 1978 Received by Power Engineering November 3, 1978 Reviewed by Power Engineering Received by Plant February 8, 1979 Comments not resolved as of last date of inspection The Request for Review and Comments (No. 212) listed no test deficiencies.
Power Engineering comments were directed toward whether; the test results met the acceptance criteria; the acceptance criteria were accurate; the test was run at the correct power level; the test prerequisites were met; deficiency reports were filed; and the use of one rod worth measurement versus the average of twr as required by the procedure were justified.
In addition, incorrect calculations in the procedure were noted by Power Engineering.
The following chronology of events were constructed from information contained in the control copy of TP 800.20 and Power Engineering comments on the test:
December 6, 1978 Request for Review and Comments sent to Power Engineering (comments requested by December 21, 1978)
December 7, 1978 Received by Power Engineering January 10, 1979 Reviewed by Power Engineering January 22, 1979 Received by Plant February 8, 1979 Comments not resolved as of last date of inspection The Request for Review and Comments (No. 218) listed i.o test deficiencies.
Power Engineering comments were directed toward; 2286 J36
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the acceptance criteria used in the testing; FSAR commitments regarding the measurement of differencial boron worths being satisfied; satisfaction of commitments made to the NRC regarding testing to be performed at 100% power; whether rod worth movements had been properly documented; and whether Technical Specification 4.10.1.2 was adhered to.
Regarding the question about conformance to the FSAR Abstract and test acceptance criteria, a memo was written to the vendor by Power Engineering on January 12, 1979.
In addition, the Project Engineer informed the Plant Superintendent that the subject testing was not in compliance with commitments made to the NRC in a letter (Serial No. 443), dated June 8,1978, in that testing was not performed at 100% power. The Plant Superintendent responded that the procedure was not intended to be run at 100% power by itself but was intended to complement the performance of TP 800.05 at 90% power. The Plant Superintendent stated that the commitment to perform the test at 100% power had been satisfied. The issue regarding compliance with commitments made to the NRC was dircussed with the Project Inspector and he will review this matter. This matter is an Unresolved Item (50-346/79-04-03).
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Administrative Procedure, AD 1801.01.2, " Conduct of Test",
states in Section 8, " Review of Test Results":
"8.1 Results of Pre-Operational and Acceptance Tests shall be initially reviewed as soon as possible, by the Test Leader, the Test Program Manager and the Section Head to determine that the acceptance criteria was met and that the procedure was executed as written.
If either one or both of these condi-tions have not been satisfied the discrepancy shall be resolved according to Section 7.0 of AD 1301.04, Resolution of Test Deficiencies. The Test Program Manager and the Section Head will initial and date the original copy of the Test Summary Report.
8.2 If the test was completed and conducted according to the approved procedure, the responsible Section Head shall distribute copies of necessary data to groups or individuals whose review is required to complete post test analysis. A Request for Review and Comment Form (Enclosure 4) shall be used for transmitting a copy of the test-related documents for review. A copy of this form shall be in the Master File. The Test Program Manager is respon-sible to see that the above reviews are completed expeditiously."
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When TP 800.05 was reviewed by the Test Leader, the Test Program Manager and the Section Head to deterutne that the acceptance criteria were met and that the procedure was executed as written the discrepancy regarding satisfying test prerequisites was not resolved in that a Deficiency Report was not issued.
When TP 800.20 was reviewed by the Test Leader, the Test Program Manager and Section Head to determine that the acceptance criteria was not met and that the procedure was executed as written the discrepancy regarding the measurement of dif ferential boron worth as required by the FSAR Test Abstract was not resolved in that a Deficiency Report was not issued.
Technical Specification 6.8.1.c requires that procedures be followed.
Failure to follow AD 1801.04 by not filing Deficiency Reports for TP 800.05 and TP 800.20 is considered to be an example of an Item of Noncompliance (50-346/79-04-01)of the Infraction level.
Regarding the timeliness of the licensee's review of test resultg a previous inspection report"jand resolution of comments, identified a concern regarding the Test Program Manager assuring expeditious review of test results by all involved parties The inspector notes that the Power Engineering comments on TP 800.05 were at the site for approximately 100 days without formal resolutior, and that the test package had not been forwarded to the Safety Review Board for review. This is discussed further in Paragraph 9.
c.
Moderator Temperature Coefficient Based on the inspector's review of the test performance and data, it appears that a positive moderator temperature co-efficient may have existed at 95% power.
A calculatiog of the power-temperature coefficient (APower/ATave) for the 8 F Tave change yields a value of +0.75% Power / F.
Therefore, this calculation taken by itself would indicate that Technical Specification 3.1.1.3 which requires a g degator temperature coefficient less positive than 0.0X10
/k F whenever thermal power is > 95% of rated thermal power was exceeded. This matter is an Unresolved Item (50-346/79-04-04) pending further review.
-2/ Inspection Report 50-346/78-17, Paragraph 2.
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5.
Core Thermal Power Evaluation The inspector examined information relating to the January 4, 1979, calculation of core thermal power as described in ST 5030.01,
" Reactor Protection System Daily Heat Balance Check".
The informa-tion included manual calculations such as Enclosure 2, " Approximate Primary System Heat Balance", Enclosure 3, " Accurate Primary System Heat Balance", Enclosure 4, " Secondary System Heat Balance", and Enclosure 5, " Weighted Primary and Secondary Heat Balance", and results celculated by the online computer.
The inspector noted that the initial conditions satisfied the requirements specified in the procedure and that the instrumenta-tion was calibrated. The inspector noted that the readings of temperatures and pressure at 5-minute intervals were taken at various points in the primary and secondary systems. These read-ings were properly recorded in the manual calculati(ns and in the computer output. The inspector verified that the computer lookup of enthalpy as a function of temperature and pressure was correct.
The acceptance criteria stated that the reactor power calculated by Enclosures 3, 4, and 5 of ST 5030.01 should agree within + 2% of the computer calculated value.
In addition, at 100% power the discrepancy between the primary (Enclosure 3) and the secondary (Enclosure 4) heat balances should be less than + 1%.
The inspector verified all manual calculations. The percentages of full thermal power calculated in Enclosures 3, 4, and 5 and from the computer were 98.4%, 99.4%, 99.4%, and 99.1% respectively. The inspector concluded that the acceptance criteria were met.
The inspector noted that incorrect entries on soce of the calculation sheets were struck out, but not dated and initialed as required by Administrative Procedure AD 1801.01, Section 7.3.5.
The licensee stated that the procedure would be more closely followed in the future. No items of noncompliance or deviations were identified.
6.
Control Rod Worth Measurements The inspector examined information relating to the July 23, 1978, measurements of control rod worth at hot zero power performed as described in ST 5010.03.1, Enclosure 11, " Rod Worth Data Analysis".
The inspector verified that the reactimeter was installed and operational, and the initial conditions prior to the measurements were within the range appropriate for the measurements.
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inspector noted that the measurements were performed by inserting Control Rod Assemblies (CRA) Groups 7, 6, 5, and 4 while simultane-ously diluting the reactor coolant system. The changes of the CRA positions and their related changes of reactivity were calculated by the reactimeter and plotted on strip charts.
The differential rod worths were obtained from the measured reactivity worths versus the changes in CRA positions.
The integral worth of each CRA was obtained by summing its differential worths.
The' acceptance criteria stated that for CRA Groups 5, 6, and 7 l (predicted worth measured worth)/ measured worth *1001< 15, and for the total of CRA Groups 4 5, 6, and 7, l (predicted worth measured
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worth /measuredworth*100(<10. The results of the measurements indicated that the acceptance criteria were met.
The inspector noted that there was no provision in the testing procedure to verify that the input paramaters (delayed neutron fractions and decay constants) were inputted correctly into the reactimeter. The licensee stated that such a provision would be considered in the future.
No items of noncompliance or deviations were identified.
7.
Shutdown Margin Determination The inspector examined information relating to shutdown margin determination as described in System Procedure SP 1103.15,
" Reactivity Galance Calculation".
g Technical Specifications require that the shutdown margin always be greater than 1% of reactivity. The inspector noted that there were two methods for shutdown margin determination.
For an average reactor coolant system temperature less than 500 F, the control boron concentration should be larger than the predicted value that incorporated the 1% of reactivity for shutdown margin.
For an average temperature greater than 500 F the shutdown margin should be calculated from the worths of fuel, boron concentration, and control rods.
The inspector examined a calculation performed on September 14, 1978, for 90% power, an average temperature of 582 F, a boron concentration of 1125 ppmB, and a burnup of 103.9 effective full power days. The inspector reviewed the time sharing computer program that calculated the worths of xenon and samarium, verified the input and output of the computer run for this calculation, and 2286 ;40
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o examined the reactor operator's curve book which included the plots of fuel worth, boron worth, and boron correction factor as functions of burnup. The inspector noted that the control rod worth was properly obtained for this calculation; however, the boron worth appeared to be incorrect because its value was taken from an inappropriate curve. The error did not affect the conservatism of this calculation and the computed shutdown margin was still greater than 3% of reactivity and satisfied the Technical Specifications.
The inspector noted that the error should not have escaped the attention of the checker and the reviewer. The licensee agreed that improvement in the checking and the reviewing of calculations was needed and indicated he would consider this matter. No items of noncompliance or deviations were identified.
8.
Unresolved Items Unresolved items are matters about which more inf ormation is required to ascertain whether they are acceptable items, Items of Noncompliance, or Deviations, Unresolved Items disclosed during the inspection are discussed in Paragraphs 3 and 4.
9.
Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on February 9, 1979. The inspectors summarized the purpose and scope of the inspection and the findings. The following items were discussed:
a.
The status of previous inspection findings and unresolved items.
b.
Power coefficient of reactivity determination and rod worth measurements. The licensee stated he would resolve the comments on the subject tests (TP 800.05 and TP 800.20) by February 16, 1979.
c.
Process computer software. The licensee stated he would furnish documentation by March 1, 1979 as to whether the computer software had been submitted to NRR for review.
d.
High pressure injection issue. The licensee stated he would inform the inspector by telephone on February 12,1979, as to the status of this matter.
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