ML20211A578
| ML20211A578 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 05/30/1986 |
| From: | Heitner K Office of Nuclear Reactor Regulation |
| To: | Walker R PUBLIC SERVICE CO. OF COLORADO |
| Shared Package | |
| ML20211A585 | List: |
| References | |
| NUDOCS 8606110215 | |
| Download: ML20211A578 (87) | |
Text
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NUCLEAR REGULATORY COMMISSION O-
.j WASHINGTON, D. C. 20555 May 30, 1986 f
Docket No. 50-267 i
Mr. R. F. Walker, President Public Service Company of Colorado P.O. Box 840 Denver, Colorado 80201-0840
Dear Mr. Walker:
SUBJECT:
NRC COMMENTS ON THE FINAL DRAFT OF THE FORT ST. VRAIN (FSV)
UPGRADED TECHNICAL SPECIFICATIONS (TS)
Enclosed are our comments on the final draft of the FSV TS, as submitted in your letters dated November 27,1985(P-85448)andDecember 24,1985(P-85481).
We request that you submit a formal application to amend the FSV TS within 90 days of the date of this letter, per item 7.b of Enclosure 8 of the letter dated February 7, 1986, authorizing 35% power operation for FSV. Your' submittal must be fully responsive to all the regulatory requirements, especially with respect to 10 CFR Section 50.91. The changes made to the FSV TS must be completely supported by appropriate safety evaluations.
Your submittal should provide approp-f ate references to supporting analyses in your FSAR or other submittals. provides the status of our review of your response to NRC Action Item guidance provided to you by letter dated October 22, 1985. Enclosure 2 provides the status of our review of your responses to Public Service Company of Colorado (PSC) Action Items. Enclosure 3 provides our comments on your revised final draft TS submittal. As part of your submittal, you should address all of the enclosed NRC coments and list which of tne NRC-proposed resolutions are acceptable to you.
For the remaining NRC-proposed resolutions, you should offer alternative resolutions with accompanying) justifications for each. Enclosure 4 forwards a note from NRC (T.L. King to PSC (M. Holmes),
dated May 21, 1985, containing additional comments that apparently were not addressed in your earlier submittal. provides the NRC's markup of your final draft TS.
The following TS sections /LCO's are still under staff review and, therefore, were not addressed in the enclosures:
3/4.2.4, " Core Inlet Orifice Valves / Minimum Helium Flow" - ongoing review outside the Technical S,)ecification Upgrade Program (TSUP),
3/4.3.1, " Plant Protective System" - ongoing review outside the TSUP, 3/4.3.2, " Monitoring Instrumentation - Chlorine Detection and Alarm System" - ongoing review outside the TSUP, 0606110215 060530 PDH ADOCK 0D000267 P
PDR t
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m May 30, 1986 3
3.5.1.1/3.5.1.2, " Safe Shutdown Cooling Systems,"
3/4.6.2, " Reactor Plant Cooling Water /PCRV Liner Cooling System,"
3/4.6.3, " Reactor Plant Cooling Water /PCRV Liner Cooling System Temperatures,"
3.7.1.5/3.7.1.6, " Safety Valves," - staff review in progress on PSC's redrafted LCO's submitted February 28, 1986, 3/4.10.1, " Xenon Stability" - ongoing review outside the TSUP.
Due to the substantial effort involved in this review, the original date (January 30,1986) for the NRC review completion could not be met.
This delay should not affect the overall completion date schedule, since the original date for overall completion (January 1, 1987) was extended three months to April 1, 1987.
The information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.
Should you have any questions regarding this letter and the enclosed comments, pleasecontactmeat(301)492-8288.
Sincerely, original signed by Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
Enclosures:
1.
Status of PSC Response to NRC Action Item Responses 2.
Status of PSC Action Item Responses 3.
NRC Comments on the PSC November 30, 1985 Draft Technical Specifications 4.
T.L. King Memorandum to M. Holmes 5.
NRC Final Draft Markup cc w/ enclosures:
See next page DISTRIBUTION:
N' LMarsh, RSB
- To receive Enclosures 1-5.
NRC*PDR*"
ACRS(10)
All others receive 1-4 Local PDR*
RIreland, RIV DCS CHinson BGrimes OELD SSPD Reading EJordan NSIC PNoonan XHeitner JPartlow Olynch H8erkow
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Elantz, RSB
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For the following:
3.5.1.1/3.5.1.2, " Safe Shutdown Cooling Systems,"
3/4.6.2, " Reactor Plant Cooling Water /PCRV Liner Cooling System,"
3/4.6.3, " Reactor Plant Cooling Water /PCRV Liner Cooling System Tempera tures,"
3.7.1.5/3.7.1.6, " Safety Valves,"
staff review in progress on PSC's redrafted LC0's submitted February 28, 1986, 3/4.10.1, " Xenon Stability" - ongoing review outside the TSUP.
Due to the substantial effort involved in this review, the original date (January 30,1986) for the NRC review completion could not be met.
This delay should not affect the overall schedule for completion, since the original date for completion (January 1,1987) has been extended three months to April 1,1987.
The information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.
Should you have any questions regarding this letter and the enclosed comments, pleasecontactmeat(301)492-8288.
Sincerely,
%X.W Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B
Enclosures:
1.
Status of PSC Response to NRC Action Item Responses 2.
Status of PSC Action Item Responses 3.
NRC Coments on the PSC November 30, 1985 Draft Technical Specifications 4.
T.L. King Memorandum to M. Holmes 5.
NRC Final Draft Markup cc w/ enclosures:
See next page
Q O
Mr. R. F. Walker Public Service Company of Colorado
- Fort St. Vrain cc Mr. D. W. Warembourg, Manager Albert J. Hazle, Director l
Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein,14/159A Mr. J. W. Gahm, Manager GA Technologies, Inc.
Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. L. W. Singleton, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Fort St. Vrain Nuclear Station Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. 0. Box 640 Platteville, Colorado 80651 Kelley, Stansfield & 0'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 l
Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80203
^
ENCLOSURE 1 STATUS OF PUBLIC SERVICE COMPANY OF COLORADO (PCS)
RESPONSE TO FUCLEAR REGULATORY COMMISSION (NRC) ACTION ITEM RESPONSES Docket No. 50-267 E. J. Butcher Letter to 0. R. Lee Dated October 22, 1985 Public Service Company of Colorado (PSC) stated in P-85448 (p. 2 of cover letter) in reference to the Technical Specifications Upgrade Program (TSUP) that:
"NRC comments regarding the NRC action items that resulted from the July 1985 meetings between PSC and the NRC (NRC letter, Butcher to Lee, dated 11/22/85, G-85433) have been included." The summary below is the NRC evaluation of the status of how these items have been included. Actions are categorized below as they were in the original NRC letter (G-85433).
Definitions of the status terms are as follows:
Open Means either that a PSC response is ongoing or that, if the response is incorporated in the TSUP, PSC should state specifically where the information was included (e.g., page number and/or specification number in the Technical Specifications (Tech. Spec.)).
Under Review Means either that the NRC review of the PSC response is ongoing or that NRC is providing a counterproposal.
Tentatively Means that, as of the November 1985 draft NRC agrees Closed that the item is tentatively closed (final closure pending approval of the subsequent requested amendment change).
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I General Actions:
Open 2., 3.
4 Under Review 1., 5.
Tentatively Closed 4., 6.
Specific Actions:
4 Open i
Under Review 2.,
3., S.,
6.,
7.,
8.,
9.,
10., 12., 14.
Tentatively Closed 1.,
4.,
11., 13.
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I Additional Comments:
Open 4.
Under Review 1.
Tentatively Closed 2., 3.
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r 2
1.
ENCLOSURE 2 STATUS Of PUBLIC SERVICE COMPANY Of COLORADO (PSC) ACTION ITEM RESPONSES (Attachment 3 to P-85448)
The summary below is the NRC evaluation of the status of the subject items.
Definitions of the status terns are as follows:
Open Means the PSC response is ongoing Under Review
- eans either that the N#C review of the response is ongoing or that NRC is providing a counterproposal Tentatively Means that, as of the November 1985 draft, WRC agrees Closed the item is tentatively closed (final closure pending the approval of the subsequent requested amendment change.)
Open 2), 6), 9), 10), 11)a), 29), 34), 38), 59)
Under Review 4),15),18), 20), 21), 23), 25), 27), 28), 30), 31),
32), 33), 35), 36), 42), 43), 49), 50), 51), 52), 53),
55), 56)
Closed 1), 3), 5), 7), 8), 11)b), c), d), 12), 13), 14), 16),
17), 19), 22), 24), 26), 37), 39), 40), 41), 43), 44),
45), 46), 47), 54), 57), 58), 60) 6 3
o ENCLOSURE 3 1
NUCLEAR REGULATORY COMMISSION (NRC) COMMENTS ON THE PUBLIC SERVICE COMPANY OF COLORADO (PSC) NOVEMBER 30, 1985 DRAFT TECHNICAL SPECIFICATIONS NRC COMMENTS - GENERAL 1.
PSC has not followed the format of the STS regarding having all 3/4 section bases in a separate section with their page numbers labeled with a "B" to distinguish them as bases.
2.
Apparently, PSC intends to have a deleted Section 7 gap between Section 6 and Section 8.
3.
PSC uses the term "immediate" for the ACTION time response in certain LCOs.
Per NRR memorandum 8. W. Sheron to R. Fraley (ACRS) dated September 13, 1984, "immediate" is to be interpreted to mean "within ten minutes."
4.
Per previous agreement with PSC (0.R. Lee letter to H.N. Berkow, dated November 27, 1984), surveillance frequency "R," "at least once per REFUELING CYCLE" should be "at least once per REFUELING CYCLE but not to exceed 18 months."
If certain surveillances are appropriate only on a frequency of "at least once per REFUELING CYCLE", PSC should justify those-on a case basis.
5.
Satisfaction of single failure criteria whenever the CALCULATED BULK CORE TEMPERATURE is greater than 760*F is a new licensing basis that has not been reviewed / approved by the hRC and appears to be outside the original scope of the TSUP.
6.
PSC did not respond to T. L. King note to M. Holmes (see Enclosure 4).
PROPOSED RESOLUTIONS 1.
PSC should use the STS format regarding bases sections.
2.
PSC should renumber Section 8 and make it Section 7 so as to eliminate any gap in the numbering sequence.
3.
PSC should define "immediate" or change it where it appears in ACTION time responses to "within 10 minutes."
4.
PSC needs to clarify surveillances with periodicity "R" to mean either "at least once per REFUELING CYCLE but not to exceed 18 months" or "at least once per REFUELING CYCLE."
5.
The licensee should submit the analysis supporting this new licensing concept for NRC review / approval.
6.
PSC should respond to the subject note.
4
NRC COMMENT - INDEX The final draft index is not in the STS format.
PROPOSED RESOLUTION PSC should revise the index to be consistent with the STS format.
NRC COMMENTS-DEFINITION 1.11 Without further definition, the phrase "except for control rod pairs" allows too much flexibility for reactivity changes.
PROPOSED RESOLUTION PSC should further restrict "except for control rod pairs" so that undesirable positive reactivity changes are eliminated.
NRC COMMENTS-DEFINITION 1.14 First draft comments requesting definition of " Bypass Flow" and
" Describe calculational method" have not been satisfactorily resolved.
PROPOSED RESOLUTION PSC should define " bypass flow" and describe more definitively how CORE AVERAGE TEMPERATURE is calculated during STARTUP, LOW POWER, and POWER OPERATION.
NRC COMMENTS-DEFINITION 1.16 E-BAR should include radionuclides other than the noble gases.
PSC's earlier reply (P-85448 Attachment 2) still does not adequately address other isotopes, such as S-35 and H-3, in the FSV primary coolant.
PROPOSED RESOLUTION PSC should make the definition of E-BAR consistent with that of STS Rev. 5.
NRC COMMENTS-DEFINITION 1.17 l
Deviations from STS wording that are only for convenience or preference are at cross purposes with overall consistency of format.
5
PROPOSED RESOLUTION NRC final draft markup will replace " Specification 4.0.2" with
" Table 1.2" and move the appropriate intervals from Specification 4.0.2 to a new Table 1.2.
NRC COMMENTS-DEFINITION 1.22 PSC added wording which is redundant to the existing wording.
PROPOSED RESOLUTION NRC final draf t aarkup will delete the last sentence in this definition- " Nonessential Portions... is maintained."
NRC COMMENTS-DEFINITION 1.24 Deviations from STS wording that are only for convenience or preference are at cross purposes with overall consistency of format.
PROPOSED RESOLUTION
" Table 1.0-1" should be relabled " Table 1.1" in this definition and in the table itself.
NRC COMMENTS-DEFINITION PSC (P-85448 Attachment 2) did not add a definition for " RESPONSE TIME," stating it is a backfit issue.
PSC should clarify how any of their safety analyses that involve trip actuation on a plant variable have any validity without consideration of response time to calculate overshoot and variable changes from the time the trip signal value is reached until the trip actuation occurs.
PROPOSED RESOLUTION Response time of the plant protective system trip instrumentation (LCO 3.3.1) is required for instrument operability and should, therefore, be added to the definitions, to Section 3.3.1, as well as to the limiting safety system settings of Table 2.2.1-1 and the associated surveillances.
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NRC COMMENT-DEFINITION The STS requires a definition for " SLAVE RELAY TEST."
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PROPOSED RESOLUTION PSC should add a definition for " SLAVE RELAY TEST" and provide the l
associated tests.
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.o NRC COMMENTS-TABLE 1.0-1 Note at bottom of page was added to the final draft without justification.
PROPOSED RESOLUTION The NRC final draft markup will propose corrections.
PSC should also provide justification to demonstrate that automatic protective actions, alarms, etc., will not be bypassed by ISS switch setting changes, or they should reference the LCO that authorizes the bypass conditions.
NRC COMMENTS - SL 2.1.1 l
1.
In the markup of the April 1985 draft, NRC requested the references for Figures 2.1.1-1 and 2.1.1-2 (the latter now incorporated as Figure 3. ?.6-1 in Specification 3.2.6).
PSC states that these figures are not tvailable in references but that PSC may expand the updated FSAR to include the figures at some unspecified future date (Attachment 2 to P-85448).
2.
In the BASIS (page 2-4), PSC's response to the NRC-requested definition of " region radial power peaking factors (RPF)" is in error. The RPF sh'ould be equal to the average region power " density" (Preg) divided by the average core power " density" (P). The word
" density" and the bar over the abbreviations to designate average would be consistent with both the definition for intraregion power.
peaking factor (as given later in the BASIS) and with the abbreviation convention used in the original " BASIS for Specification LC0 4.1.2."
Therefore, Pcol and reg should also be used for abbreviations in P
the BASIS. As currently defined without using the term density, the ratio of Pr to P would not make sense as having a value of 1.83.
Further, in the same paragraph, NRC had previously questioned the arbitrary sounding use of the verb "was assumed." PSC has not changed that verb but merely changed the words following to refer to a " range between 1.83 and 0.4" for the values of the RPF assumed in the analysis of the limiting power-to-flow ratio.
-3.
The BASIS for Specification SL 2.1.1 states that a " conservative estimate of the most unfavorable axial power distribution was also used." An axial power density ratio is defined as "less than or equal to 0.90 plus or minus 0.09 for regions with control rod pairs fully inserted or withdrawn, and 1.23 plus or minus 0.12 for regions with j
control rod pairs inserted more than 2 feet." The actual values used in the supporting analysis are not specified.
Page 4.1-5 of the 7
O e
O P
original technical specifications, and page 5-11 of the current draft of DESIGN FEATURE 5.3.4.c imply that there is.already a 10%
uncertainty factor applied to this ratio.
The staff's interpretation is that analysis has shown that 0.99 and 1.35 are acceptable axial peaking factors but that the operating configurations for control rod pairs are selectively limited to those for which calculations produce axial peaking factors less than 0.9 or 1.23 respectively, i.e., a 10%
conservatism.
4.
In the BASIS (page 2-4), the reference to Specification 3.2.2.a.1 should be 3.2.2.a.l.a.
The reference to FSAR Table 3.5-1 should be deleted and replaced with one to Specification 3.2.2.a.2 and Figure 3.2.2-1.
The last sentence of the second paragraph should be written for clarity as follows:
"The 95/. confidence interval on experimental data was used in the most conservative manner to determine the rate of migration of the fuel kernel as a function of the fuel kernel temperature profile."
5.
ACTION a. of 24 hrs to be SHUTDOWN is inconsistent with the STS ACTION of I hour.
There appears to be no justification for being in violation of a SAFETY LIMIT for 24 hrs. PSC has also changed the specification and/or ACTION of the existing FSV LC0 3.1 without justification.
PR0p0 SED RESOLUTIONS 1.
An FSAR revision in the near future is recommended as an appropriate level of documentation for both the integral and transient power-to-flow ratio limits.
Can PSC provide a reference to an appropriate topical report correlating kernel migration with particle AT and Fort St. Vrain power-to-flow transients?
2.
In the NRC final draft markup, the third sentence of the first paragraph of page 2-4 will be replaced as follows:
"The region power peaking factor (RPF) is defined to be equal to the ratio of the region average power density ( reg) divided by the core average power density (P).
Conservative values for the calculated RPFs were used consistent with the allowable limits specified in DESIGN FEATURE 5.3.4.c for a core average outlet temperature above 1250*F."
Note that this proposed revision assumes that the analysis is based on the most conservative, or limiting, values of the calculated RPFs as implied for other peaking factors; however, without referenced i
documentation, the staff cannot be sure that conservative values have always been used in any of the analyses. PSC needs to clarify under what circumstances an RPF of 0.4 is more conservative than one of 1.83.
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3.
PSC should revise the BASIS and DESIGN FEATURE 5.3.4.c to clarify the exact limit and to explain what is being assumed in the safety analyses.
Since power peaking in the core bottom is adverse, the staff notes that the highest values should be assumed conservatively in the analysis.
PSC should specify analysis values in the FSAR and the BASIS.
4.
The NRC final draft markup will include the designated corrections.
5.
PSC should justify why they have changed from the existing FSV LC0 3.1. The PSC justification in Attachment 2 (P-85448) addresses only those changes made since the April 85 draft, not those from the existing FSV LC0 3.1.
NRC COMMENTS - SL 2.2.1 1.
In Table 2.2.1-1, PSC is inconsistent in specifying the TRIP SETPOINTS both compared to the STS and compared among the different LIMITING SAFETY SYSTEM SETTINGS in the final draft. As discussed in to P-85448, PSC argues that inserting a "less than or equal te" sign before the trip settings for items 2(c), 2(d), and 2(e) is iretorrect because a " TRIP SETPOINT" is not an absolute value and should not have a limiting value.
Instead, PSC specifies the trip setting plus or minus a tolerance which apparently equals the " margin for drift" as discussed on page 2-16 of the BASIS. However, PSC does not use the same format for the trip settings in items 1(a) through 2(b) in Table 2.2.1-1.
For Item 1(c) in Table 2.2.1-1, the Upper Trip Setpoint and the Upper Limit (text for ALLOWABLE VALUE) should also indicate applicability "for circulator inlet temperature equal to or greater than 742*F."
NRC had requested a reference to a figure showing the TRIP SETPOINT versus temperature, but PSC indicated that figure 2.2.1-1 for the ALLOWABLE VALUE versus temperature was the only reference available.
PSC gives no justification as to why they should not be required to provide a reference. Also, why is psia used instead of psig?. Are indication and calibration both in psia?
2.
In Table 2.2.1-1, what are the TRIP SETPOINT uncertainties on primary coolant moisture dewpoint temperature instrument settings?
3.
In Figure 2.2.1-1, the staff questions the reasons for using psia versus psig. The note on the figure for Allowable High Primary Coolant Pressure should be rewritten to make the linear equation more obvious, e.g., PRESS. - 0.6338*(TEMP) + 276.7, in one continuous line.
4.
In the BASIS for SL 2.2.1, the General Methodology subsection is provided to expand upon NRC comments for definition of " allowable value." What is missing is a table in either the BASIS or the FSAR comparing " allowable values" to " analysis values" so that any reviewer can readily ascertain the degree of conservatism being used in the FSAR.
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5.
On page 2-17 of the BASIS, the reference for the nuclear detector decalibration methodology.is not yet documented in the updated FSAR.
j 6.
In the BASIS under the subsection entitled, Primary Coolant Moisture-Hich, an explanation is needed as to why the TRIP SETPOINT and the ALLOWABLE LIMIT are the same with no uncertainty band on the TRIP SETPOINT.
PROPOSED RESOLUTIONS 1.
The NRC final draft markup will include suggested changes.
PSC should provide FSAR references for the TRIP SETPOINTS and ALLOWABLE LIMITS and should be consistent in how the TRIP SETPOINTS are presented in the Tech Spec.
PSC needs to address the issue of usage for psia versu; psig.
2.
PSC needs to define instrument uncertainties for primary coolant moisture dewpoint temperature monitors.
3.
PSC needs to revise Figure 2.2.1-1 as indicated on final draft markup and to resolve dual usage of psia versus psig.
4.
Prior to final draft approval, PSC needs to provide a table in the FSAR to compare ANALYSIS VALUES versus ALLOWABLE VALUES for each accident analyzed.
5.
Prior to final draft approval, PSC needs to revise the FSAR to document the detector decalibration methodology.
6.
Same as 2 above.
NRC COMMENTS - LCO 3.0 1.
Item 3.0.3.
The time limit interval clarification for the ACTION l
Statement in the STS Rev. 5, P. 3/4.0-1 starting with "where corrective measures..." is missing.
2.
Item 3.0.4.
The added words "or as required by automatic or manual protection ACTION" should only result in passage to reduced power modes or SHUTDOWN, and as such are redundant to the existing words.
3.
Item 4.0.2.
The footnotes are not consistent with the STS intent in spite of the PSC discussion (P-85448 Attachment 2).
It is the express intent of the STS not to allow doubling of intervals, even infrequently.
4.
BASIS Items 3.0.3 and 4.0.6 should be updated to reflect any changes made in these sections.
+
5.
As worded, there is no requirement for NRC review / approval or the use of ASME,Section XI, Division 2 as appropriate.
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PROPOSED RESOLUTION 1.
NRC final draft markup will add the time limit clarification "where corrective measures..." to Item 3.0.3 as on P. 3/4.0-1 of STS Rev. 5.
2.
NRC final draft markup will delete the added words "or as required by automatic or manual protective action."
3.
NRC final draft markup will delete the footnotes in Item 4.0.2.
4.
NRC final draft markup will change BASIS Items 3.0.3 and 4.0.6 in accordance with the proposed resolutions above.
PSC should ensure that all subsequent sections of the technical specifications that reference the renumbered 4.0 surveillances are correctly renumbered.
5.
NRC markup of final draft will propose a new Section 6.17, " Inservice Inspection and Testing Program."
NRC COMMENTS - LCO 3.1.1 1.
In LC0 3.1.1, SR 4.1.1 and the accompanying BASIS, the use of the terms " scram time" in relation to a value of "152 seconds" should be clarified since, in every case, the terminology refers to an
" extrapolated scram time" of "152 seconds from the fully withdrawn position." As currently written, the LC0 or SR could be misinterpreted to refer to a scram time less than from the fully withdrawn position and, thereby, lead to an allowance larger than intended by either PSC or the staff. The staff notes that the
" extrapolated scram time" is to be determined from the partial scram test (SR 4.1.1.c), but clarification is needed as to whether PSC intends also to qualify the partial scram test for ascertaining control rod " operability" by the "back-EMF" method independent of the
" extrapolated scram time" estimation.
Finally, Table 3.2-2 in the updated FSAR, Revision 3, shows a scram time of 160 seconds, maximum.
2.
LC0 3.1.1 lacks a control rod pair operability condition on moisture levels in the helium purge flew. An ACTION statement is needed to ensure that tne dry helium source is connected when needed to preserve control rod operability and t6 ensure that surveillances are performed to establish the operability of the reserve shutdown system per SR 4.1.6.
There is also a need to provide sufficient defense-in-depth to reduce or eliminate sources of moisture ingress into the helium purge flow.
Such considerations are addressed in NRC Comment 2 to LC0 3.6.1.3.
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SR 4.1.1 lacks CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST l
surveillance requirements for the helium purge line knockout pot, moisture, element, and pressure transmitter.
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LC0 3.1.1 lacks a control rod pair operability condition on primary coolant impurity levels consistent with the graphite and burnable l
poison corrosion limits specified in LC0 3.4.2 and LC0 3.4.3.
Because of the possible connection between observed high primary coolant moisture levels (about 100 ppm) prior to the partial failure to scram reported in R0 50-267/84-008, an ACTION is needed at least to increase the frequency of the partial scram test (SR 4.1.1.c).
The frequency should be increased from once per 7 days to a frequency consistent with the increasing severity of primary coolant impurity levels and the accompanying allowable time periods for continued operation with out SHUTDOWN as specified in the ACTION statements for LC0 3.4.2 and 3.4.3.
The staff judges that increasing the frequency of the partial scram tests under conditions of high primary coolant impurity concentrations will allow the early detection of any degradation of control rod operability.
Fo'r example, LCO 3.4.2 ACTION Statement a allows 10 days of operation when the total impurity level (H 0, 2
CO, C0) exceeds 10 ppm but is less than 100 ppm.
During the 2
allowed period of continued operation under these conditions, increasing the frequency of the partial scram test from once per 7 days to once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> would appear to be an appropriate response to assure early detection of CRDM degradation.
5.
LC0 3.1.1 ACTION Statement a needs clarification with regard to the use of the word "immediately" as recommended by Section 4.1(1) of American National Standard ANSI /ANS-58.4-1979. The staff suggests "within 10 minutes" as a realistic alternative to "immediately."
Also, in ACTION Statement a, the determination / confirmation of SHUTDOWN MARGIN per Specification LC0 3.1.3/SR 4.1.3 should be required when one or more control rods are " immovable."
6.
In the final draft, LC0 3.1.1 ACTION Statement d indirectly imposes the requirement to perform partial scram tests at an increased frequency per SR 4.1.1.b.1.a.2 and 3.
However, LCO 3.1.1 ACTION Statements e and f do not specifically require the performance of a partial scram test following loss of purge flow. An ACTION Statement to perform the partial scram test should be included for all situations in which CRDM conditions are altered, even temporarily. A j
l reasonable but short time limit (such as "within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />") should also be specified for performing the partial scram test.
7.
LC0 3.1.1 ACTION Statement d.2 should read "215'F" rather than "250*F" to be consistent with SR 4.1.1.b.l.a.2.
In fact', to make the LCO fully consistent with the SR, ACTION d.2 should read as follows:
" Surveillance testing per Specification 4.1.1.b.l.a is performed on the control rod pair (s) once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the CRD motor temperature exceeds 215*F."
l 8.
SR 4.1.1.b.1.a.2 should cite Specification 4.1.1.c not 4.1.1.b with regard to performing the partial scram test.
l l
12 l
l
9.
SR 4.1.1.b.2 does not specify a minimum purge flow as previously requested by NRC.
- 10. The BASIS indicates that the control rod pair withdrawal accident analyses described in FSAR Sections 1 4.2.2.6 and 14.2.2.7 were performed assuming a scram insertion time of 152 seconds and a ramp reactivity insertion of 0.080 delta K.
Further the position is implied that the 0.080 delta K assumed in the accident analysis is readily achieved because the total calculated reactivity worth of all 37 control rod pairs is 0.210 delta K which is significantly larger than the assumed 0.08 delta K scram reactivity. These statements raise several points which require some clarification.
First, Table 3.2-2 of the updated FSAR, Revision 3, gives the maximum rod insertion time for scram as 160 seconds not 152 seconds.
Section 14.2.2.3 of the FSAR cites this table for the scram time assumed in the accident analysis.
I Second, Table 14.2-2 of the updated FSAR gives minimum scram reactivities which vary from 0.064 to 0.093 delta K, depending upon initial condition; the 0.08 delta K cited in the BASIS is not given as a typical example in Table 14.2-2, nor is it quoted anywhere in the text of Section 14.2.
The value quoted in the BASIS appears to be based on that quoted in Section 3.5.3.1 (page 3.5-5 in reference to Table 3.5-6), but there is no clear connection between these values and those used in the safety analyses.
Third, the 0.210 delta K given as total worth of the 37 control rod pairs is not truly representative of the total available scram worth.
For example, at hot, equilibrium full power conditions out to at least MOC, there must be' sufficient rods inserted to compensate for 0.02 delta K in core excess reactivity (Table 3.5-4, updated FSAR).
Approximately 0.08 delta K is estimated for the total burnup reactivity needed to achieve 300 EFPD. This estimate derives from combining the c.02 delta K in equilibrium full power excess reactivity with 0.06 delte K available from the depletion of burnable poisons
(
(Table 3.5-1, updated FSAR).
This estimate, whicn is deduced solely
)
from the FSAR, may also be a little low since the report GA-A12795, dated March 1974, quotes a burnup reactivity of 0.053 delta K for six months (or 180 EFDH) in Cycle 1 which linearly extrapolates to 0.088 delta K at 300 EFPD. Section 3.5.1 of the FSAR states that Cycle 1 was based on an exposure of 150 EFPD (or six months at 83 percent load factor) which extrapolates to 0.106 delta K.
The Base Reactivity Curve for Cycle 4 transmitted by PSC letter P-85261, dated July 24, 1985, implies a burnup reactivity of 0.11 delta K at 180 EFPD, but the composition and use of this curve is not adequately covered in the BASIS for LC0 3.1.7/SR 4.1'.7.
The interpretation of this curve is therefore in doubt.
Under the most likely circumstances, the best estimate for core excess reactivity during much of core life runs between 0.02 and 0.04 delta K, thereby reducing the available scram worth at equilibrium full 13
)
power from 0.21 to between 0.17 and 0.19 delta K.
In this situation also, three or more control rod pairs may be inserted fully with up to four others inserted deep into the core to provide for control maneuvers (see Table 3.5-2 of the FSAR for Cycle 1 rod sequencing).
In this configuration, the scram reactivity rate for 37 rods (as given in Figure 3.5-2, which is cited for accident analysis in Section 14.2.2.3) may not be representative of the expecteo scram response but is apparently assumed to be bounding or conservative, i.e., higher worth at a slower rate of insertion.
In fact, for LOW POWER and STARTUP, in which rod pair insertion is greater than that of i
equilibrium full power conditions in order to compensate for the loss in negative reactivities due to Xenon-135 and temperature defect (up to 0.032 and 0.044 delta K respectively per Table 3.5-4 in the updated FSAR), the available scram reactivity is again reduced, perhaps substantially.
Figure 3.5-2 is much less applicable for describing the actual reactivity insertion rate.
The FSAR analysis has attempted to use a " bounding assumption" for scram analysis in terms of slower rate of insertion. While at first glance it appears to be bounding, no separate effects analysis is provided to show that it is indeed bounding.
In summary, the BASIS needs to be reconciled with the FSAR, and the discussion of the FSAR assumptions needs to be enhanced.
- Further, since available scram worth is directly tied to core excess reactivity, which in turn is determined by fuel and burnable poison loadings, PSC should at least expand on the importance of burnable poisons both in DESIGN FEATURE 5.3.4 and in the BASIS for LCO 3.1.7/SR 4.1.7.
11.
In the BASIS under Temperature Limitation, additional clarification is needed on automatic operation and alarm functions associated with the monitoring of CRDM motor temperatures.
References are needed.
- 12. The significance of the 272*F limit mentioned in the BASIS needs further clarification.
Does the 272*F limit apply to the CRDM or to the CRDM insulation? Why is there no automatic or manual scram required when CRDM temperatures reach the 272*F limit discussed in the BASIS? An LC0 ACTION Statement similar to ACTION Statement a is appropriate for the 272*F limit.
References are also needed as to the sources of the data on the CROM temperature limits.
The staff notes that Section 3.8.1.1.2 (page 3.8-8) of the updated FSAR states that "the maximum temperature rating of the CRDM is 272*F (Ref. 8)."
Should not the FSAR reference be cited here also?
- 13. The determination of CRD M motor temperature by comparison (SR 4.1.1.b.1.b) requires the delineation of applicable uncertainties, particularly for those determinations involving the "other factors" discussed in the BASIS.
PSC should document their uncertainty analysis and describe how the uncertainties are ured.
14
14.
The BASIS discussion of Purge Flow requires amplification and clarification of the uses and functioning of the knockout pot, moisture element, and pressure transmitter.
For example, does the knockout pot have a second alarm at 6 inches of condensed water?
Specific references to the updated FSAR and other descriptive documentation are required.
3 15.
The BASIS discussion of Action and Surveillances requires further information, references, and clarification of:
the required operator response to water ingress into the purge line; the establishment of measurable minimum purge flow rates for detection purposes; and the operator response to ensure control pair operability during periods when moisture and other oxidants exist at high levels within the primary coolant.
16.
There is no SR in response to slack cable alarm (Section 7.2.2.2, updated FSAR).
The BASIS does not discuss the slack cable alarm as being indicative of an "imnovable" control rod pair.
The absence of slack cable indication or alarm should be given as one of the i
conditions under which control rod pairs are considered OPERABLE in LC0 3.1.1.
l PROPOSED RESOLUTIONS 1.
The NRC final draft markup will incorporate necessary changes.
PSC needs to clarify the intended uses of the partial scram test including any future use of the "back-EMF" method.
2./4./6.
The NRC final draft markup has provided the LCOs on moisture and impurity levels.
PSC should provide the other changes and/or additions.
3.
PSC needs to propose the required SR 4.1.1 CHANNEL CALIBRATIONS and CHANNEL FUNCTIONAL TESTS for purge line moisture instrumentation.
5./8.
The NRC final draft markup will incorporate necessary changes.
7.
PSC should make the ACTION and Surveillance consistent relative to 215'F or 250*F.
l 9.
As concurred in by NRC in July 1985, PSC will specify a minimum purge flow at a later date when a sufficient operating history has been obtained (see Attachment 2 to P-85448).
10./11./12.
PSC should provide needed clarification, changes, and/or additions.
13./14./15./16.
PSC should provide needed clarification, changes, and/or additions.
9 15
NRC COMMENTS - LCO 3.1.2.1 1.
In SR 4.1.2.1.a.3, the words " incapable of being with drawn" are used in reference to certain fully inserted control rod pairs.
This terminology is not defined in the BASIS for this LCO but is defined on page 3/4.1-34 of the final draft of the BASIS for LCO 3.1.4.2 and SR 4.1.4.2.
The definition of " control rod pairs incapable of being withdrawn" should be included in Section 1.0, DEFINITIONS, and either repeated or so cited in the BASIS of this LCO.
2.
In SR 4.1.2.1.a.3.b and in SR 4.1.2.1.b.2 the words "less than or equal to 6 inches" should be replaced with " greater than 0 inches but less than or equal to 6 inches."
3.
In SR 4.1.2.1.c.2, the words "less than 6 inches" should be replaced with " greater than 0 inches but less than 6 inches of withdrawal."
4.
In SR 4.1.2.1.c.2, should the words "If the analog and digital position indications indicate" be rewritten "If the analog or digital position indication indicates?" If the indications disagree, does the operator believe the digital as implied in SR 4.1.2.1.a.3.a. unless the analog is shown to be accurate by other means? Clarification is needed.
5.
In the first paragraph of the BASIS, PSC should clarify whether the control rod pair " predicted" position refers to one that is
" calculated" or " calculated plus a calculation-to-measurement bias."
In the ORNL Monthly Report to NRC-AE00 for August 1985, PSC is quoted as indicating that the " Cycle 4 Base Reactivity Curve" (submitted in PSC letter P-85261, dated July 24, 1985) includes a reactivity bias to the calculational model used to generate the curve.
The implications and significance of any reactivity bias to the expected critical control rod pair positions and shutdown margins should be explained with regard to assessing the adequacy of the calculational models.
Since comparisons of calculated ano measured data on each cycle's performance are not submitted in reload topical reports for FSV as is done for LWRs, such data could be included periodically in FSAR revisions to establish the base line for reactor physics operating data.
6.
In the BASIS, PSC should give the reasons for assuming that the digital control rod pair position is more accurate than the analog when the two indications disagree.
7.
In Table 3.1.2-1 and the BASIS (page 3/4.1-14) reference is made to the " watt-meter test" as an independent means to establish control rod position.
In the BASIS, PSC should elaborate upon how this test is performed and provide appropriate references to test descriptions and NRC concurrence in its usage (e.g., SER dated April 22, 1985).
This terminology is not used in either Sections 3.2.2.6, 3.8.1.1.1, 7.2.2, or Appendix A.9 of the updated FSAR, although page 3.8-5 describes preinstallation tests in which drive motor vattage was recorded and correlated with conditions preventing scram.
Further clarification is needed on the details and status of the " watt-meter test."
16
8.
The BASIS states that "a 10 inch position accuracy for all control rod pairs is also consistent with a reactivity uncertainty of about 0.003 delta K, which allows for detecting core irregularities." PSC should elaborate on this statement with regard to effects on control of power distribution and the determination of core reactiv*ty.
Section 3.5.2.2 (page 3.5-5) of the FSAR states that the average worth of a control rod pair is about 0.005 delta K.
Table 3.2-2 indicate; that the nominal full rod stroke is 191.3 inches.
It appears that the 0.003 delta K reactivity uncertainty can translate to a 60% or 115 inches uncertainty in control rod pair position if position indication is lost. A discussion of backup mechanisms to prevent an unidentified loss of position indication is appropriate, including slack cable indication and alarms, outlet thermocouple indication, and any other mechanisms which are available to, routinely checked by, or alarmed for, the operator.
The BASIS also states that Section 7.2.2 of the FSAR assumes a long-term misalignment of 12 inches, from which the BASIS infers an allowable 2-foot separation distance between rod pairs in a group; however, Section 7.2.2.1 indicates only that there should be an alarm on " excessive deviation between rod pairs in a group."
Section 7.2.2.2 indicates the alarm actuates for deviations greater "than 2 i i ft. with respect to each other." Therefore, the FSAR allows deviations between 12 and 36 inches, not the 24 inches implied in the BASIS.
PROPOSED RESOLUTIONS 1.
PSC should provide a DEFINITION using words similar to those used on page 3/4.1-34.
2./3.
The NRC final draft markup'will include the suggested addition and corrections.
- 4. through 8.
Prior to approval of the final draft, PSC should reconcile the BASIS with the FSAR and should provide the NRC with necessary clarifications, additions, and/or other changes.
NRC COMMENTS - LCO 3.1.2.2 1.
In SR 4.1.2.2.a.3, a reference should be made to a DEFINITION for
" control rod pairs incapable of being withdrawn."
2.
In SR 4.1.2.2.d.2, the words "less than 6 inches" should be replaced with "areater than 0 inches but less than 6 inche?, of withdrawal."
3.
In SR 4.1.2.2.d.2, PSC should clarify whether the words "If the analog and digital instrumentation indicate 6 or more inches" should be rewritten as "If the analog or digital instrumentation indicates 6 or more inches."
17
1 i
4.
The BASIS cites experimental data on control rod pair worth versus with drawal positions as differing from older calculations but as substantiated in newer analyses.
References are appropriate for these measured and calculated data since available reports (e.g., GA-A14007, dated February 1977) provide no information on such discrepancies from early reactor testing, and since no such discussion accompanies figure 3.5-2 as cited in Section 3.5.3.1 of the updated FSAR.
Do these data change the FSAR assumptions relative to accident analyses in Section 14.27 5.
The BASIS gives a range of bank position uncertainties from 17 inches at fully inserted to 13 inches at fully withdrawn.
These estimates are derived from the combination of relative bank worth versus withdrawal position (such as illustrated in Figure 3.5-2, updated FSAR), a total bank worth (37 rod pairs) of 0.21 delta K, and a reactivity uncertainty of 0.003 delta K.
The discussion in the BASIS is appropriate for possible mispositioning or position uncertainty of the 37-rod bank, but, as discussed in Comment 8 to LC0 3.1.2.1 above, PSC should address the safety significance of single-rod position uncertainties, which may be as great as 115 inches for a 0.003 delta K reactivity uncertainty.
Also, Section 7.2.2.2 of the FSAR indicates that deviations within a group may be as large as 36 inches before the operator receives an alarm.
6.
The BASIS cites control pair withdrawal procedures without citing a specific reference.
7.
The BASIS uses the term " predicted position" with out defining whether
" predicted" means " calculated" or " calculated plus bias."
If the latter, the source and nature of the bias assumption should be documented in the BASIS or in a reference.
8.
The BASIS (page 3/4.1-18) uses the terminology " control rod pairs capable of being withdrawn." This terminology requires a DEFINITION to ensure clarity as to the technical and administrative requirements.
The BASIS makes a reference to Specification 3.1.4.2 as listing "the methods of 'aaking a control rod pair incapable of being withdrawn," but LC0 3.1.4.2 does not specifically give that definition, although the definition intended does appear in the BASIS for LCO 3.1.4.2 (page 3/4.1-34).
See Comment 1 above.
9.
The BASIS (page 3/4.1-18) uses the term "This Surveillance" without specific reference to an GR.
- 10. The BASIS needs to elaborate on and to provide reference citations for the " watt-meter test" which is cited in the ACTION Statements.
See Comment 7 on LC0 3.1.2.1 above for further clarification of needs.
11.
In PSC letter P-85242, dated July 10, 1985, the Interim Technical Specifications for Reactivity Control included two figures of scram worth versus withdrawal distance on pages 3/4.1-18 and -19 of the BASIS for LC0 3.1.3./SR 4.1.3, the predecessor of the BASIS for LCO 18
3.1.2.2/SR 4.1.2.2 in the final draft. The BASIS did not discuss these figures. Why have the figures been deleted and why was their significance / utility (if any) not discussed in the BASIS?
12.
The requirements of SR 4.1.2.2.d do not agree with the BASIS on page 3/4.1-19.
13.
In the ACTION Statement, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> are allowed for Actions a, b, or c, in the SHUTDOWN mode.
The STS (Section 3.1.3) requires immediate opening of the reactor trip system breakers.
PROPOSED RESOLUTIONS 1.
PSC should include and reference a clarifying DEFINITION in Section 1.0.
The words for this DEFINITION can be based on that given on page 3/4.1-34 of the final draft.
2.
The NRC final draft markup will incorporate the suggested changes.
3.
PSC should provide clarification.
4.
PSC should provide references and any necessary changes to the BASIS and the FSAR.
5.
PSC should provide additional clarification on the significance of rod position uncertainties as related to reactivity uncertainties.
- 6. through 10.
PSC should provide the required clarifications, additions, references, and/or changes, 11.
PSC should provide clarification as to the uses of the deleted figures.
12.
The NRC final draft markup will propose a correction.
13.
PSC should change their ACTION for SHUTOOWN to agree with the STS or justify why not.
NRC COMMENTS - LCO 3.1.3 1.
In reviewing LCO 3.1.3 and SR 4.1.3, the staff notes that these specifications and their BASES are very unclear in describing how the operator is to determine compliance with the SHUTDOWN MARGIN limit.
Furthermore, the staff's interpretation is that PSC essentially relies on calculations to determine compliance with this LC0 and to perform the SR.
The staff requires further clarification as to the status of qualification and verification of such calculations.
Comments 5 and 10 below address the staff's concern about how the operator is able to assess the SHUTDOWN MARGIN in an effective and timely manner.
Comment 4 reflects the staff's current understanding of the status of PSC's SHUTOOWN MARGIN calculational models with respect to quality assurance and verification.
In addition, most of the Comments on LC0 19
9 3.1.7 are applicable here also with regard to quality assurance in the calculation models and methods.
The Specifications and supporting operational techniques must provide the operator with a simple and readily verifiable indication of reactor shutdown or of the capability to achieve shutdown when the plant is operating.
Based on available information from the BASIS and FSAR, it is very unclear as to whether such reliable indication is available to the operator.
2.
The staff notes that ACTION Statement a.1 was apparently written to be consistent with the approach and format of the LCO 3.1.1.1 ACTION Statement in the W-STS.
However, PWRs have two normally available reactivity control mechanisms, i.e., normal boration/dcooration from the chemical volume control system (CVCS) and control rod insertion / withdrawal.
There is also a reserve shutdown mechanism through boration from the diverse safety injection systems, some portions of which are shared with the CVCS. The PWR can be borated by the normal control function from the CVCS, thereby reducing power and increasing the instantaneous SHUTDOWN MARGIN per W-STS DEFINITION 1.25 because the available negative reactivity for assuring HOT SHUTDOWN via control rod scram is unaffected.
The required action at Fort St. Vrain should be more consistent with that of LC0 3.3.1 ACTION Statement a. in the GE(BWR)-STS since HTGRs and BWRs have only one normal control / shutdown mechanism, i.e.,
control rods, plus a reserve shutdown mechanism.
The staff's suggested rewording of LC0 3.1.3 ACTION Statement a.1 is consistent with the approach used in the GE(BWR)-STS. Also, the 24-hour time limit in LCO 3.1.3 ACTION Statement a.2 should be reduced to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to be consistent with LC0 3.1.1 ACTION Statement a for immovable control rod pairs, LCO 3.1.2.1 Table 3.1.2.1 ACTION 1 for loss of control rod pair position indication, and LC0 3.1.7 ACTION Statement for the difference between observed and expected core excess reactivity exceeding 0.01 delta K.
The cited ACTIONS in LC0 3.1.1, LCO 3.1.2.1, and LC0 3.1.7 are in response to observable conditions which can directly impact the available SHUTOOWN MARGIN, and 50 the response times should be the same.
It is also noted that, when LC0 3.1.3 ACTION Statement a.2 is pursued with the 12-hour limit as recommer.ded, then ACTION b should ensure that the 0.01 delta K SHUT 00WN MARGIN is effected within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> following discovery that the "available" SHUTOOWN MARGIN is less than 0.01 delta K.
3.
LCO 3.1.3 ACTION Statement c.1 should be revised to remove the ambiguity in the use of the word "Immediately" per the guidance of section 4.l(1) of American National Standard ANSI /ANS-58.4-1979.
4.
SR 4.1.3 begins with the statement that " SHUTDOWN MARGIN shall be verified as follows:".
SR 4.1.3.a.2 follows and begins with the words that "In assessino the SHUTOOWN MARGIN the following conditions shall be assumed:".
The section on SHUTDOWN MARGIN - OPERATING in the BASIS (page 3/4.1-24) implies that " verification" is provided by an
" assessment" which relies solely on " calculations and analysis" not directly on " test and experiments."
Does " verification" literally 20
f mean " calculation"?
If so, where is the traceable path of verification documentation for the calculations of subcritical multiplication in FSV?
The staff notes that the GA critical experiments (as reported in GA-8468, dated February 1968, and Gulf-GA-A9354, dated January 1973, and as cited in Section 3.5.7 of the updated FSAR) did not include measurements of subtritical multiplication.
In the GA experiments, the critical multiplication was established only in nonuniform configurations for which large reactivity corrections for material substitutions had to be ertimated in order to infer " excess reactivities" for both the rodded and unrodded " uniform" core configurations.
The large magnitude of the reactivity corrections in these experiments does not appear to be well correlated with the small uncertainties quoted.
From the limited comparison of calculated and measured data (as illustrated in Table 3.5-22 of the FSAR), there appears to be at least a 0.013 delta K discrepancy between rodded and unrodded calculations.
The uncertainties in these " critical" experiments have not been reviewed, may be too small, and may not be applicable to the analysis of subcritical multiplication. Also, the calculational methods and data used in available analysis of these experiments do not appear to be comparable to those used for FSV.
The pulsed-neutron measurements of FSV subcritical multiplication and subcritical control rod worth from FSV Startup Test A-3 (reported in GA-A13079, dated July 1974 and cited in Section 13.3.3 of the updated FSAR) appear to be well correlated with calculations (Tables 6-1 and 6-3, GA-A14007, February 1977), although specifics and details are generally lacking. Those experiment-to-calculation correlations included the use of the "better" or more exact BUGTRI/ GAMBLE models, which are apparently not used in following FSV cycle reactivity with burnup. However, there are also reported correlations to a two-dimensional GAUGE model, and GAUGE is one of the tools used to calculate FSV performance with burnup. As indicated in GA-A14007, the GAUGE model homogenized the control rod pair over the fuel block and used " adjusted or corrected" rod densities to force agreement with BUGTRI/ GAMBLE results and/or experimental data. The report (GA-A14007) fails to specify the number of neutron energy few groups used in the GAUGE model.
As noted in the ORNL monthly report to NRC-AE00 for August 1985, there are two GAUGE models, one using four groups and another using seven groups. As noted in the ORNL report, each model has its own reactivity biases hot and cold and with burnup.
It is not clear from the BASIS which model is used to calculate the SHUTDOWN MARGIN or even if a totally different calculational tool is used.
For quality assurance, the SHUTDOWN MARGIN analytical model should have been calibrated against the Cycle 1 BOC (cold) BUGTRI/ GAMBLE results for the subcritical pulsed-neutron experiments, and then periodically compared against the "better" BUGTRI/ GAMBLE results for different cycles, different burnups in cycle life, and different core temperatures and Xenon-135 conditions.
Such comparisons would be necessary to assess any variance, bias, or other uncertainties which 21
might have developed in the calculation of subcritical multiplication using the SHUTDOWN MARGIN analytical model.
In addition, the effect of any noted bias or other difference between calculated and measured critical multiplication and rod worths on the FSV core (such as derived from SR 4.1.4.2 and SR 4.1.5) should be factored correctly into the assessment of the model used to calculate SHUTDOWN MARGIN.
Results of these comparisons should be available in documented form.
Without the documentation of the quality assurance and verification efforts for the calculational models used to assess SHUTDOWN MARGIN, the terminology " verification / verified" cannot be used interchangeably with words such as " calculation / prediction / calculated / predicted" and.
" assessment / assessed," as done in Specification 3/4.1.3.
PSC needs to clarify and carefully define its word usage with respect to
" verification."
5.
SR 4.1.3.a raises questions with regard to the operator's ability to ascertain compliance with LC0 3.1.3.
SR 4.1.3.a implies that once per 7 days either the operator or a support staff engineer determines the core burnup to date and then either performs a core calculation at that burnup or an interpolation off,a precalculated tabulation of SHUTDOWN MARGIN versus burnup.
The key question derives from the required assumption for assessing the SHUTDOWN MARGIN (as given in SR 4.1.3.a.2.b) and regarding the statement "with all inoperable control rod pairs in their pre-scram positions." For a weekly surveillance, the staff surmises that the operator can use a precalculated tabulation which incorporates the assumption of one, two, or even three " worst" stuck fully withdrawn rod pairs. What is not clear is what mechanism the operator has available if an arbitrarily large number of rod pairs are found immovable while performing SR 4.1.1.c or fail to scram, as occurred in R0 50-267/84-008 in June 1984.
In the latter instance, the operator had to wait a day to get the results of a GAUGE calculation.
If LCO 3.1.1 ACTION a applies, the operator has 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to be in at least SHUTDOWN.
But can he ascertain whether LCO 3.1.3 ACTION a also applies within the allowed I hour, or does he have to wait 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be sure, as in the June 1984 event? Table 3.5-8 of the updated FSAR gives the average rod pair worth for each three rod pair group.
Is the operator provided with a crude but conservative tabulation of individual rod worths from which he can estimate the loss in SHUTDOWN MARGIN for any arbitrary set of stuck rods? Clarification is needed.
6.
In SR 4.1.3.b.1, does the statement "when all control rod pairs cannot be verified fully inserted" mean the same as "when any one or more control rod pairs cannot be verified to be fully inserted"?
7.
InSR4.1.3.b.2,doesthek.tatement"ifallcontrolrodpairsarenot fully inser ted prior to withdrawal action" mean the same as "if any one or more control rod pairs are not fully inserted prior to withdrawal action"?
i 22
.?
8.
In SR 4.1.3.b.4 and SR 4.1.3.c
.3, does " assessing" mean " verifying" as ir. plied in the opening statement of SR 4.1.37 If so, justification is needed as noted in comment 3 above.
9.
In SR 4.1.3.c.3.a. a reference appears needed to a DEFINITION for
" control rod pairs capable of being withdrawn." A similar comment applies to SR 4.1.3.c.3.c with regard to rod pairs " incapable of being 1
withdrawn." Both need a DEFINITION.
10.
In SR 4.1.3.c.3.b. a situation is implied in which the " control rod pairs" are to be " withdrawn for... SHUTDOWN MARGIN verification,"
after the SHUTOOWN MARGIN has supposedly already been assessed in order to verify the SHUTDOWN MARGIN per the opening statement of SR 4.1.3.
Why does PSC in tend to perform rod pair withdrawal for SHUTDOWN MARGIN " verification" after already performing " verification" by assessment of calculated results? Clarification is needed as to what constitutes " verification." The staff recognizes that fresh fuel is potentially involved in performing SR 4.1.3.c but questions how rod withdrawal can provide SHUTDOWN MARGIN verification.
The calculated worth of the withdrawn rod pair may exceed 0.01 delta K, but that calculation in turn has to have a documented, traceable path of verification by comparison to other experiments or tests.
The staff notes that any changes in subcritical multiplication are not correlatable with the FSV startup channel count rate because of the highly nonlinear inverse multiplication of these detectors during subcriticality and the approach to criticality (Section V.2, EPRI-NP-698, February 1978 and pages 22-23, Trans. Am. Nucl. Soc. 26.
Supplement, August 1977).
The BASIS needs to be clarified as to what information is gained by the rod withdrawal and how such information is used in a quantitatively or quantifiably meaningful way (other than qualitatively assuring that the core appears to remain subcritical when a " calculated" amount of reactivity is inserted) to
" assess / verify" the SHUIDOWN MARGIN.
The staff's current interpretation is that PSC relies on calculations basically only for quantifying the SHUTOOWN MARGIN. However, those calculations are not demonstrably verified.
PROPOSED RESOLUTIONS 1./2./3.
The NRC final draft markup will include some suggested changes; however, PSC also needs to provide the indicated clarifications.
4.
PSC needs to provide clarifications and references.
5.
PSC needs to provide clarifications in the BASIS.
6./7.
The NRC final draft markup will include suggested changes.
8.
PSC needs to provide similar clarification and references as also noted in Comment 3.
9.
PSC needs to provide an appropriate DEFINITION.
10.
PSC needs to provide clarifications.
23
NRC COMMENTS - LC0 3.1.4.1 1.
The title of Specification 3/4.1.4 (page 3/4.1-27) should reverse the words " WORTH" and " POSITION" to be consistent with the first sentence of LC0 3.1.4.1.
2.
LC0 3.1.4.1.a and b does not have a specific limiting condition associated with the control rod pair withdrawal sequence.
In this respect, LC0 3.1.4.1.a and b differ significantly from the original LC0 4.1.3 (from which it is derived) as well as from the original upgrade draft, dated April 1, 1985.
Section 7.2.2.1 (item 10) of the updated FSAR cites the control rod pair withdrawal sequence interlock as being " required to assure rods are withdrawn in a prearranged sequence." Further, the same citation states that " excessive deviation between rod pairs in a group has to be alarmed."
In Section 7.2.2.2, the alarm function is stated to actuate for within-group deviation of 2 i i ft.
The BASIS makes reference to DESIGN FEATURE 5.3.4 as providing the specification of a control rod withdrawal sequence.
This is not the case, since DESIGN FEATURE 5.3.4 only specifies allowable peaking factors for various rod positions, not sequences., The BASIS implies that meeting the peaking limits equates to complying with the required withdrawal sequencing, but that connection is at best tenuous without a surveillance on what is actually programmed into the interlock.
SR 4.1.4.1.2 requires verifying BOC group worths against calculations based on the assumed withdrawal sequence, but SR 4.1.4.1.1 does not require verification of the withdrawal sequence itself.
DESIGN FEATURE 5.3.6 acknowledges that "the distribution of fuel of different ages requires the use of alternatr-shim rods each year of full power operation," but this also does nut translate into a surveillance requirement on rod pair with drawal sequence.
In summary, there is no actual LCO on control rod pair withdrawal sequence, only indirect allusions to an action that the FSAR describes as being " required."
The LCO, SR, and BASIS require revision to incorporate the FSAR requirements on ensuring the withdrawal in a prearranged sequence, with a fixed within-group deviation allowed.
3.
LC0 3.1.4.1.c.1 limits the control worth of an individual rod pair to less than 0.047 delta K (given in Section 14.2.2.7 of the FSAR and greater than the 0.043 delta K estimate in Section 14.2.2 or the 0.03 delta K estimate in Section 3.5.3.1) for the just critical source power core; however, the source power level is specified as "E-07" and is missing a number, presumably 1.0, before the E.
4.
LC0 3.1.4.1.c.2 cites a reference (the AEC SER dated January 20, 1972) which is a little short on' details.
The AEC SER (page 33-34) encompasses all the reactivity insertion accidents in Sections 14.2.2.6, 14.2.2.7 and 14.2.2.8 of the FSAR.
The specific transients include a reactivity insertion up to 0.025 delta K at full power, (0.012 delta K is later cited as the reference case at full power), 0.047 delta K at source power, and a 37-rod pair bank position 24
B reactivity insertion rate of 0.0029 delta K per second at either full power or source power.
These values are more specific than the indirect reference to the AEC SER, which states that any such transients should either be terminated at 140 percent power with peak fuel temperature below 1500*C or on a trip at a hot reheat steam temperature of 1075'F, with less than 2 percent fuel failures and no loss of system integrity.
The LCO should be rewritten around actual worth limits which are presumed to be more readily calculated and then confirmed by measurements on the core.
5.
SR 4.1.4.1.1 is readily performed for determining compliance with LC0 3.1.4.1.a and b as currently written in the final draft, but SR 4.1.4.1.2 is apparently meant to be ascertained only from BOC rod pair and group worth measurements.
As currently written, LC0 3.1.4.1 is very weak.
However, LCO 3.1.7 implies that the fundamental mechanisms already exist to provide a means for operators to determine compliance with both the position and worth requirements of LC0 3.1.4.1.
In effect, SR 4.1.7 provides for direct surveillance on control rod pair positions and worths once every 7 days.
For example, the operator must already be able to establish core burnup from the integral of the plant's thermal power output over time. SR 4.1.7.a and the BASIS for LCO 3.1.7/SR 4.1.7 imply that the operators do this to determine core burnup for interpolation of a
" base reactivity curve."
If burnup is so determinable, then the operator should also be able to use a set of bounding expected critical position curves calculated for equilibrium burnups at given constant power levels, and also for hot and cold restarts.
The curves could be provided both with and without extra rods inserted less than two feet for power peaking control.
Given the estimated burnup and closest equilibrium condition obtaining in the core, the operator should be able to ascertain whether the control rod pair positions and, by inference, the rod pairs worths are within acceptable uncertainty limits from the calculated critical position curves.
Deviation could be noted and assessed against the individual rod pair worth requirements in the controlling rods (LCO 3.1.4.1), the SHUTDOWN MARGIN (LC0 3.1.3), the temperature defect (LC0 3.1.5), and the overall core reactivity status (LC0 3.1.7).
If such curves are not to be used, the operator can still assess compliance with LC0 3.1.4.1 every 7 days when he performs the normalizations required to compare against the " base reactivity curve" per SR 4.1.7.a.
This second option lacks timeliness but p.ovides a weekly check on positions and worths.
PSC should revise LC0 3.1.4.1 to compare measured rod pair positions against calculated or predicted positions, and thereby infer deviations in rod pair worths.
This approach has precedent, as described in the BASIS for Specification 3/4.1.2 of the GE(BWR)-STS. Usa of position curves 25
?
versus burnup is highly recommended to be consistent with the operator response times imposed in the ACTIONS for LC0 3.1.4.1.
Performing the surveillance every 7 days in conjunction with SR 4.1.7 would not be appropriate if the ACTION times are only 12 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
6.
In the BASIS, the reference to the AEC SER (page 3.4.1-29) should be replaced or supplemented with the pertinent references to FSAR sections.
7.
In the BASIS (page 3/4.1-29), the " maximum" control rod pair worth cited for the control rod pair withdrawal accident analysis in the FSAR is quoted at 0.012 delta K.
Section 14.2.2.6 cites the 0.012 delta K as used in the " reference" rod withdrawal analysis.
Section 14.2.1.1 states that this corresponds to " worst conditions" at E00. However, the maximum value used in the safety analysis appears to be 0.025 delta K (as also noted in Section 14.2.2.6), which is larger even than the 0.016 delta K BOC maximum discussed in Section 14.2.1.1.
Clarification is needed as to what " maximum" means in this context.
8.
The BASIS (page 3/4.1-30) makes several references to the power peaking factors discussed in DESIGN FEATURE 5.3.4.
The staff notes that it is difficult to correlate the peaking factors with their uncertainties and significance to safety analysis.
Sections 3.5.4.1 through 3.5.4.3 of the FSAR are generally correlated with the peaking factors quoted in DESIGN FEATURE 3.5.4, in terms of finding an equivalent number quoted. However, the constituent parts of the peaking factors including uncertainties are not so readily apparent, nor is how they relate to power peaking uncertainties quoted for the thermal analysis (Section 3.6.4.b).
The staff notes that although th! BASIS for SL 2.1.1 makes reference to the 10 percent axial peaking Jncertainty, DESIGN FEATURE 3.5.4 has dropped the uncertainty values tnat appeared in the BASIS of the l
original LC0 4.1.3.
The peaking factor uncertainties quoted in the original LCO 4.1.3 may derive from answers to licensing questions III.1, III.2, and III.3 on the FSV PSAR.
These answers are documented in Attachment A to Amendment No. 2 to the FSV PSAR (00CKET-50267-4, May 31, 1967) and Attachment A to Amendment No. 3 to the FSV PSAR (DOCKET-50267-13, July 1967) with partial revisions in Attachment A to Amendment No. 9 to the FSV PSAR (00CKET-50267-10, January 4, 1968).
The peaking factor uncertainty estimates given in the PSAR amendments were derived from a statistically nonrigorous assessment of the Peach l
Bottom Critical Experiment Correlation (primarily GA-3799, March 1963; other documents cited were never submitted to the AEC-DRL) and the Peach' Bottom Unit 1 Startup Test (GAMD-7354, October 1966).
Because the nuclear data and methods used for these older, and not very representative, analyses differ from those used at FSV and because the l
fuel lattices are significantly different, the uncertainty estimates l
are questionable. More information is needed, both on the calculation of the peaking factor and on the origin and justification of uncertainties. Why did PSC drop the reference to uncertainty factors in the peaking factor limits quoted in DESIGN FEATURE 5.3.47 1
26 l
?
s 9.
In the BASIS (page 3/4.1-30), the parenthetical reference to FSAR Section 3.5.4.3 should be to 3.5.4.2.
l
- 10. The BASIS (page 3/4.1-31) maintains that the measurement of control rod pair group worth at BOC provides adequate comparisons to calculations. SR 4.1.4.1.2 indicates that reactivity worth is measured at BOC only for "the control rod pair groups from LOW POWER to POWER in the withdrawal sequence." By comparing Tables 3.5-4 and 3.5-6 to Table 3.5-8 of the FSAR, it is estimated that, at least i
for Cycle 1 BOC, there could be only about three rod groups subjected to such measurements between criticality at the refueling tempcrature (220'F) and that at the operating temperature (1500*F).
This estimate may be high if LOW POWER measurements start at temperatures higher than the refueling temperature (220*F).
As illustrated in Table 3.5-8, one of the Cycle 1 BOC control rod pair groups withdrawn is' a very low worth group (less than 0.001 delta K).
From Table 3.5-4, the smaller temperature defect and larger excess reactivities in equilibrium cycles imply even less control rod motion for performing such rod group worth measurements.
Power range rod worth measurements are genera-lly recognized to be more complicated and less accurate (i.e., greater uncertainties) on most reactors including HTGRs. The BASIS should explicitly acknowledge that group worth measurements are limited to, at most, a quarter of the total bank worth, and probably much less; the BASIS should incorporate appropriate uncertainties for extrapolations to the total bank worth.
If power range measurements are being included (another two and a half groups withdrawn), a discussion of power range measurement uncertainty should be added. The assumed 20% uncertainty is noted to be much larger than the PWR 10% uncertainty on reactivity computer inferences and 5% on borate concentration endpoints, where virtually all the bank worth is measured twice (in and out) at zero power.
However, there is no indication that the 20% uncertainty accounts for any complications beyond normal calculational and measurement erroi bands, such as including extrap'olation uncertainties on the total bank worth.
Clarification is needed as to how much of the total bank worth is actually measured and how uncertainties are applied and extrapolated.
Which rod pairs are measured, those in fresh or exposed fuel? Are nonstandard rod configurations allowed and obtained during BOC startup as in BWRs? If so, where is the requisite documentation? Does the
" withdrawal sequence" include or preclude such configurations?
PROPOSED RESOLUTIONS 1.
The NRC final draft markup will include this change.
2.
PSC should incorporate the required revisions to the LCO, SR and BASIS to ensure compliance with a specified rod pair withdrawal sequence.
- 3. through 8.
PSC needs to provide additional clarification and changes as indicated.
l 21
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9.
The NRC final draft markup will include this change.
10.
PSC needs to provide additional clarification.
NRC COMMENTS - LCO 3.1.4.2 1.
In LCO 3.1.4.2.a. the references to control rod pairs " incapable of being withdrawn" should be supported by a DEFINITION in Section 1.0 similar to that given in the BASIS. As noted in the comments to other LCOs, the terms " incapable of being withdrawn" and " capable of being withdrawn." appear often, and so, need a uJ..' mon DEFINITION.
2.
In LCO 3.1.4.2.a.1, does the statement "up to two control rod pairs may be removed from the PCRV" refer just to maintenance and refueling configurations? To accommodate malfunctions, should this statement read "any two control rod pairs may either be removed from the PCRV or immovable in a withdrawn position"? This wording adds a measure of legitimacy to the reference to "the allowable two" in ACTION Statement a.2.
Action Statement a.2 is in effect long before rod pairs could be removed from the PCRV.
3.
In LCO 3.1.4.2.a.2, the reference to " SHUTDOWN MARGIN verification" should read " SHUTDOWN MARGIN assessment." As noted previously in comments on other LCOs, the verification is indirect.
Subcriticality may be verified directly by the rod pair withdrawals, but the worth of the withdrawan rod pairs is available onlyfas a calculated value.
The calculations must be subjected to " verification" through a documented, traceable path which both compares calculations to quantitative test data and evaluates the biases and uncertainties..The BASIS should provide a brief discussion of, and reference to, the " verification" of the calculational methods. Also the words "or tests" need to be clarified or deleted.
4.
In LCO 3.1.4.2 ACTION Statements b.,
b.1, and b.2, the words "Immeaiately" should be clarified to the times involved. Also in ACTION Statement b.,
the words "above requirements" need clarification as to whether they mean ACTION Statement a. or LC0 3.1.4.2.a.
5.
In SR 4.1.4.2.b.2, the parenthetical reference to the " watt-meter test" should be supported by a reference citation or other enhanced i
description in the BASIS.
(
6.
In SR 4.1.4.2.c. the wording states that "the SHUTDOWN MARGIN shall be explicitly calculated per the requirements of Specification 3.1.3."
In LC0 3.1.3, the word " calculated" never appears.
In SR 4.1.3, the words are " verified, " followed by " assessing." In the BASIS for LCO 3.1.3/SR 4.1.3, assumptions are given which could be used in a calculational analysis, but the words " calculation" and " calculated" never appear, although " verification" appears on pages 3/4.1-24 and 3/4.1-26. Clarification is needed as to terminologies and their precise meanings.
28
o 0
7.
In SR 4.1.4.2.d, the verification of the. SHUTDOWN MARGIN must again be noted as being based on two separate verifications:
one documented for the calculations and calculational models, and the second observed for the core subcriticality. The first requires references to supporting documentation in the BASIS; the second requires some j
amplification as to exact techniques.
Is subcriticality confirmed "by demonstrating that the source range countrate is below an expected value and that the time integral of source range rate of change is zero"? The BASIS should also be clarified on this point.
8.
In the BASIS (page 3/4.1-35), the statement is made that "After shutdown, verifying that each control rod pair is fully inserted ensures that the position of each control rod pair is known and that the SHUTDOWN MARGIN calculation is accurate." Verifying that each control rod pair is fully inserted merely assures the operator that the core configuration is consistent with that which should have been simulated in the SHUTDOWN MARGIN calculation.
Tne accuracy of the SHUTDOWN MARGIN calculation is a completely separate matter which is most likely out of the reactor operator's purview.
Tne accuracy of the SHUTDOWN MARGIN calculation should be determinable from a documented quality assurance and methods verification program, which in turn should be referenceable at this point in the BASIS.
9.
In Attachment 2 to P-85448, PSC noted that SR 4.1.4.2.c was modified so that an explicit SHUTDOWN MARGIN calculation is not required for the removal of only one control rod drive assembly from the reactor; however, a SHUTDOWN MARGIN calculation is required for the removal of more than one control rod drive assembly.
PROPOSED RESOLUTIONS 1.
Proposed resolution is the same as NRC Comment 1 to LC0 3.1.2.2.
2./3./4.
The NRC final draft markup will incorporate suggested changes, but PSC also needs to provide additional clarification in the BASIS, and clarification of the words "above requirements".
5./6./7./8.
PSC needs to provide clarification.
9.
In Attachment 2 to P-85448, PSC presented the justification for the change which NRC agreed to at the July 1985 meeting; however, PSC should provide the verification documentation of the SHUTDOWN MARGIN calculations.
The FSAR should be updated accordingly.
NRC COMMENTS - LCO 3.1.5 1.
Compliance with LC0 3.1.5 is only verified at BOC per SR 4.1.5 by measurement of the temperature coefficient as a function of temperature between 220*F and 1500*F. The temperature d'.fect is obtained by integration of the measured temperature coefficient over 29
temperature.
No mechanism is provided for verifying the temperature defect as a function of cycle burn up.
Should a M0C verification be required if the core reactivity limit of LC0 3.1.7 is violated and no other short-term effect can be credited? Is the temperature coefficient measured only at Xenon-free, startup power levels? Is the value of the temperature coefficient affected by Xenon transients and l
load follow maneuvers?
2.
The limits in LC0 3.1.5 for maximum and minimum temperature defects are not found explicitly stated in Sections 3.2, 3.5 or 7.2 of the FSAR. The BASIS describes these allowable limits as having a 10%
uncertainty factor.
Are the larger and smaller analysis limits without the uncertainty factor the ones used in safety analyses?
Where are the analysis and allowable limits described in the FSAR?
Tables 3.5-4 and 3.5-6 incorporate temperature defects to evaluate core excess reactivities and shutdown margins respectively.
Typical Xenon-free temperature defects to the refueling temperature are listed in Table 3.5-4.
All but the initial BOC value fall within the limits. Why is the BOC an exception? The BASIS for the original LCO 4.1.5 states that the minimum calculated value is 0.028 delta K (the lower limit without uncertainty in the current BASIS).
If the calculation shows that the minimum will reach 0.028 delta K, how can that be prevented? By lowering the allowable operating temperature?
Clarifications are needed as to what is actually calculated, and how uncertainties are actually being applied.
The discussion in the BASIS about the derivation of the maximum temperature defect makes reference to Table 3.5-4 but uses numbers for excess reactivities and excess SHUTDOWN MARGIN not given in the FSAR? What is the " excess SHUTDOWN MARGIN" relative to? How is it compared to values in Table 3.5-67 3.
The BASIS (page 3/4.1-38) ends by stating that the " Performance of the Surveillance Requirements verifies the assumptions used in the analysis."
If the " analysis assumptions" are those used in the FSAR Section 14.2, then the FSAR needs to be updated to clearly spell out the analysis versus allowable values as noted in Comment 2 above.
If the " analysis assumptions" are those used in the neutronics calculational models, then sufficiently detailed topical reports need to be referenced, describing the methods, nuclear data and simulation models, their quality assurance program and verification against test data.
For example, how do methods and nuclear data used for FSV calculations cor., pare to those used to derive comparisons in Tables 3.5-19 and 3.5-31 of the FSAR for other HTGR lattices? There are also significant variations in the carbon-to-uranium (C/U) ratios between FSV and the configuration for which comparisons are given in Tables 3.5-19 and 3.5-31.
What are the relative contributions to the temperature defect from Doppler broadening and the thermal neutron scattering law in graphite (including the effect of different graphite impurities) in these other HTGR lattices as compared to FSV? Without this information, the applicability of comparisons cited in Section 3.5 of the FSAR to FSV is suspect.
30
^
PROPOSED RESOLUTIONS 1./2./3.
PSC needs to provide clarifications and documentation.
NRC COMMENTS - LC0 3.1.6 1.
The RSD units' operability conditions in LCO 3.1.6 lack a clear statement with regard to moisture levels.
No moisture instrumentation has been added to the RSD hopper pressurization lines from the gas bottles as can be noted on PI-11-1, Issue BJ. However, as described in the BASIS to LC0 3.1.1 and in R0 50-267/84-012-03, a moisture knockout pot, moisture sensing element, and pressure transmitter have been added to the inlet purified helium purge lines to the CRDM and refueling penetrations in which the RSD hoppers sit.
These are illustrated also on PI-11-1, Issue BJ.
The implication drawn from the corrective actions listed in R0 50-267/84-012-03 is that PSC concludes that moisture ingress to the RSD hoppers could have occurred from backflow inleakage resulting from the refueling penetration through the orificed vent, which equilibrates hopper pressure with CRDM and refueling penetration pressure during normal operation. Therefore, an operability condition for the RSD units is appropriate, based on allowable moisture levels in the CRDM purge flow and the availability of a dry helium purge source.
An ACTION Statement is also needed consistent with that proposed for LCO 3.1.1 in NRC Comment 2.
SR 4.1.6, concerning RSD functional testing following water ingress, needs to be cited in the ACTION Statement.
SR 4.1.6 should reference SR 4.1.1.b.3 on verifying that purge flow is not carrying condensed water. Also, an SR for CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST of moisture instrumentation should be made by reference to SR 4.1.1 modified as proposed in NRC Comment 3 to LC0 3.1.1.
As currently written, neither LC0 3.1.1 nor LCO 3.1.6 leads to performing SR 4.1.6.f. since there are no operability conditions nor ACTION Statements on the presence of condensed moisture in the purge line.
Both LCO 3.1.1 and LC0 3.1.6 should be modified as indicated to ensure that the response to water ingress is both proper and prompt.
2.
LC0 3.1.6 ACTION Statement a should be modified to change "available" to "available for installation" in reference to the "0PERABLE spare RSD unit."
3.
The BASIS (page 3/4.1-42) addresses several post-shutdown reactivity changes for which the RSD system has been designed to compensate.
To ensure clarity, positive or negative reactivity contributions of each change should be indicated.
4.
SR 4.1.6.e.4 does not explicitly state that the RSD assemblies to be tested must be from a CRD assembly removed from the core during the current refueling.
5.
LCO 3.1.6 ACTION a. was changed without justification from the interim tech spec regarding the fourteen days.
31
PROPOSED RESOLUTIONS 1.
PSC should provide the required changes. PSC also needs to provide clarification in the BASIS on protection against, and response to, moisture ingress.
2.
The NRC final draft markup will include the proposed change.
3.
PSC needs to provide necessary clarification.
4.
The NRC final draft markup will include the necessary change.
5.
PSC should provide justification for the change.
NRC COMMENTS - LC0 3.1.7 1.
The LCO 3.1.7 ACTION Statement should refer to meeting the ccndition of LC0 3.1.3 with respect to establishing the SHUTDOWN MARGIN whenever there is an excessive deviation in core reactivity.
2.
SR 4.1.7.a requires comparing expected reactivity, based on current core burnup at full power, to a calculated reactivity for the core
- burnup, SR 4.1.7.a states that the " reactivity must be normalized to a calcu:ste: redclivity for that core burnup," but does not elaborate on what the normalization entails, nor how " reactivity" is defined, i.e., for what condittor.s. The BASIS describes a " calculated base reactivity curve" which is inferred to account for "lce; term reactivity effects involving fuel and lumped bu'rnable poison depletion and fissi.on product poison buildup", but not to include effects of "Sm-149, Xe-135, Pa-233 and fuel temperature."
Presumably, the normalization referred to in SR 4.1.7.a is the action by the operator of correcting for actual plant conditions to the base conditions which are alluded to but not specifically defined in the BASIS, e.g., what is the temperature assumed for the base reactivity curve; is the core assumed to be isothermal; etc. The " base reactivity curve" is not defined either in the FSAR or in the BASIS of the original LCO 4.1.8, from which the draft LC0 3.1.7 is derived.
Since the " base reactivity curve" is the tool to be used to ascertain compliance with LC0 3.1.7, the FSAR should include a detailed description of the derivation and uses of the " base reactivity curve" and at least provide examples of the curves for both Cycle 1 and the assumed equilibrium cycle.
Similar data and explanations are provided to some extent for other core parameters in Section 3.5.
As noted in NRC Comment 5 to LC0 3.1.4.1, the staff observes that the steps assumed to be taken to nornelize the actual operating core conditions to the base reactivity curve would appear to involve use of the same data required for the estimation of a calculated critical position for the actual operating conditions.
Is this so? 'Could the mechanism be automated to allow the operator an on-line means for determining compliance with an expanded LC0 3.1.4.2 by comparing predicted and actual rod pair positions instantaneously?
32
3.
In PSC's letter P-85261, dated July 24, 1985, the Cycle 4 Base Reactivity Curve shows large negative values of reactivity becoming more negative with lifetime. Although not described in the BASIS, l
except to refer to " normalization to an initial base state core condition," does the base reactivity curve represent a: fully or partially rodded core configuration? Also, the reactivity adjustments cited in P-85261 were associated with a 4-group GAUGE model.
Has PSC committed to always using a giveh analytical model, or aces PSC consider itself free to change models or adjust results if necessary to preclude violating the limits'of LCO 3.1.7? These situations appear to need clarification on uses of dedicated analytical moaels, the need for documented descriptions and verification, and the exposition of model bias and uncertainty. As currently written and described in the BASIS, LC0 3.1.7 could be meaningless in terms of being a genuine limit, 4.
In Attachment 2 to P-85448, PSC refers to " discussion concerning approval of changes to the Base Reactivity curve and where in the Technical Specifications reference to the curve and its approval should be made." The resolution as given by PSC is that the BASIS specifies approval "by the NFSC prior to use," which it does.
However, ADMINISTRATIVE: CONTROL-6 S.2.7.1 (page 6-19) cites the Base Reactivity Curve, but only in terms of " review and audit," not
" approval". The expansion of Specification 6.5.2 "to include-responsibility for approvina the Base Reactivity Curve in use" has not been performed consis. tent with the PSC commitment and is, therefore, inconsistent with the BASIE.
Specification 6.5.2 should designate the NFSC as approvina the Base Reactivity Curve.
The NFSC should include one or more members wno are cognizant of reactor physics issues. An outside consultant'may be required if sufficient objectivity and expertise are not'ava11able within PSC.
5.
As noted in NRC comment 11 to LCO 3.1.1, there is no specific LCO other than LC0 3.1.7.to ensure proper loadings of fuel and burnable poisons at 80C.
From Table 3.5-1 of the updated FSAR, burnable poisons are noted to compensate for'3.06 delta K in potential core excess reactivity. The importance of LC0 3.1~.7 to reactivity determinations at STARTUP after*. loading fresh fuel should be discussed in the BASIS.
Is an LC0 also reeded on startup count rate during SHUTOOWN and REFUELING to assess changes in subcritic;11ty? Is such a mechanism feasible? Should the BASIS address,the quality assurance program checks that ensure proper fuel and burnable poison loadings?
Is such a discussion more proper in the BASIS for LCO 3.1.3/SR 4.1.37 6.
In Section 3.5.7.3 of the FSAR, the observed and predicted core reactivities of Peach Bottom Unit 1 Cores 1 and 2 are discussed briefly.
Figures 3.5-21 and 3.5-22 of the FSAR illustrate the Peach Bottom HTGR cores' reactivity as a function of burnup.
The definitions and derivations of the reactivity values given in these figures are not provided in the FSAR, nor are descriptive references provided. How do the data presented in FSAR Figures 3.5-21 and 3.5-22 1
- 33
,1*
j
relate and compare to the FSV Base Reactivity Curve? Are there similar comparisons available for FSV Cycles 1 through 3 consistent with the presentation of the Peach Bottom HTGR core reactivity data?
What significant conclusions about FSV reactivity behavior can be drawn from the comparisons of the base reactivities from the two plants? Are such comparisons factored into the assessment of FSV reactivity? If not, what are the reasons for including the Peach Bottom data in the FSAR? Why has Figure 3.5-22 not been updated (see Figure 9.6-1 in the report Gulf-GA-Al2652, dated July 1973)?
7.
If FSV and the GA critical experiment evaluations (Section 3.5.7.2 of the FSAR) use the same calculational methods and nuclear data (a question which is not specifically answered in the FSAR), then Tables 3.5-13 and 3.5-26 of the FSAR provide data which indicate that FSV analysis methods (at least for cold conditions) should overpredict the critical core multiplication by 0.0075 1 0.006 delta K at BOC of Cycle 1 and by 0.0042 1 0.0006 delta K in reload cycles.
The quoted uncertainties are the least squares sum of the calculational and measurement uncertainties, and indicate a potential for exceeding the limits of LC0 3.1.7.
Can PSC establish whether this potential is real? Also, similar comparisons of calculations and measurements from the GA critical experiments, as presented in Table 3.5-22 of the updated FSAR, indicate that, as control rod pairs are withdrawn from the FSV core (i.e., the core becomes less rodded to compensate for temperature defect or burnup), the calculated core reactivity may become more positive than the measured reactivity by perhaps up to 0.013 delta A.
These results indicate that the core reactivity status with respect to the Base Reactivity curve could exceed the limit of LC0 3.1.7 as the core is he'ated up or the cycle burns out.
Has PSC seen evidence of such reactivity discrepancies? In the ORNL monthly report to NRC-AE00 for August 1985, PSC is quoted as noting a hot-to-cold bias of this magnitude for the 4-group GAUGE model which is used to generate the Base Reactivity Curve. Was the prior discovery of this bias considered by PSC at the time to be a violation of the current LC0 4.1.8, and was the reactor shut down accordingly?
What is PSC's interpretation of " expected reactivity"? Could it be sufficiently liberal so that this LCO will never apply?
8.
In discussing FSV core reactivity in Appendix III.3.A (page III.3.A-2) to Attachment A to Amendment No. 3 of the FSV PSAR (DOCKET-50267-13, dated July 1967), problems were cited in evaluating core reactivity in high neutron leakage graphite-moderated cores (i.e., certain UCRL criticals). Were these problems solved? The PSAR reference given for the " apparent solution" does not address these problems at all.
FSV is a low leakage core, but the smaller GA HTGR criticals experienced i
l higher neutron leakage.
Are the results of the analyses of these smaller experiments relevant to FSV core reactivity behavior? Are there rod pair configurations in FSV that promote higher than normal fast neutron leakage? Is it possible that the FSV core has certain critical configurations in which there is a higher probability of obtaining a violation of LC0 3.1.77 If so, does the Base Reactivity Curve account for this? Why were these high neutron leakage experiments cited in the PSAR licensing action but not described in the FSAR as another benchmark point for FSV analysis methods?
34 1
In the report GA-A14007, dated February 1977, calculations of cold critical multiplication, subcritical rod worth and axial flux distributions were reportedly adjusted to give better agreement with FSV startup measurements merely by increasing the upper reflector natural boron equivalent impurity level from 0.5 ppm to 5 ppm.
As shown in Table 6-2 of GA-A14007, the assumed correction to reflector impurity levels was worth -0.005 delta K, or half the limit in LCO 3.1.7.
This higher impurity level is reportedly supported by subsequent analysis, out no references are cited to describe the reflector block impurity analyses.
The staff notes that increasing calculated fast neutron leakage into the rodded upper reflector would have the same effect as incorporating thermal neutron absorbing impurities into the reflector. Are the reflector impurity data so well established and documented that a calculational error in fast neutron leakage is not a likely source of error?
How would burnout of reflector impurities from high thermal neutron fluxes affect the Base Reactivity Curve ? How would an error in calculating fast neutron leakage affect the Base Reactivity Curve? If the curve is based on a two-dimensional GAUGE model, how are burnup-dependent three-dimensional effects being accounted for?
9.
Are the analytical models used for generating the Base Reactivity Curve (LCO 3.1.7), for calculating the SHUTDOWN MARGIN (LC0 3.1.3),
and for the FSVFAS (LC0 3.2.1) all the same? How do they differ in terms of core reactivity predictions? How are they consistent in terms of core reactivity predictions? Why are they not compared in the FSAR? Do these models account for burnup of graphite impurities?
Will not the burnup of graphite impurities change the trajectory of the Base Reactivity Curve (LC0 3.1.7) as well as reduce temperature coefficient (LCO 3.1.5)?
- 10. Since core excess reactivity cannot be directly observed, the word
" observed" needs to be clarified. Also " excess reactivity" needs to be clarified or defined.
PROPOSED RESOLUTIONS 1.
The NRC final draft markup will include suggested changes.
- 2. through 10.
PSC needs to provide clarifications.
ilRC COMMENTS - LC0 3.2.1 1.
LC0 3.2.1 refers to "naximum in-core irradiation." Since each fuel cycle is based on an average core burnup equivalent to 300 EFPD (measurable by integrating reactor thermal power output over time),
six fuel cycles equate to an average core burnup of 1800 EFPD, the peak limit given in LCO 3.2.1 for in-core components.
The BASIS states that "in-core irradiation lifetime... is not readily 35
determinable by plant operation," which presumably can be used to determine the average lifetime in EFPD for the fuel cycle. The BASIS 1
states, in effect, that, to ascertain compliance with this LCO, the Fort St. Vrain Fuel Accountability System (FSVFAS) must be used.
By implication, FSVFAS is a calculational tool "using the calculated data which is generated for inventory purposes." FSVFAS is not described in the FSAR, although it is the sole mechanism for determining LCO compliance as much so as any instrumentation system. Appropriate documentation is required, including verification.
2.
The LC0 3.2.1 ACTION Statement is inconsistent with SR 4.2.1.1 (should be SR 4.2.1 since there is no 4.2.1.2).
SR 4.2.1 is performed " prior to entering STARTUP following each refueling" while the LCO 3.2.1 ACTION Statement is to "be in SHUTDOWN within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />" if the limit is exceeded.
Is there no periodic surveillance during burnup which is useful and appropriate? SR 4.2.1 should be qualified in its wording by inserting "by verified calculations" after "it shall be determined." As noted in Comment 1 the BASIS should provide appropriate references to indicate that FSVFAS provides " verified calculations." Verification of fuel element calculations presumably can be handled in part by post-irradiation fuel surveillances.
PSC needs to clarify whether this is being done, and where it is documented.
3.
The original specification of an average burnup within a fuel region of 110,000 MWD per tonne (original LC0 4.1.1) has been deleted per NRC markup of the previous draft.
Does FSVFAS yield results for fuelled components in units of EFPD or MWD per tonne? Does the 1800 EFPD limit a~1 ways equate to 110,000 MWD per tonne?
4.
The BASIS cites some of the FSAR Sections in Chapter 3 which define l
the irradation limits in terms of EFPD. As discussed in Section 3.4.2 and 3.8.1.2, the actual limits used to assess mechanical performance of core components are in units of fluence.
The BASIS and the FSAR, in describing the FSVFAS, should indicate how EFPDs are correlated to fluence values, and should present the verification of such correlation mechanisms.
Fuel performance limits in MWD per tonne (FSAR Section 3.2.1) and FIMA (FSAR Appendix A.2) require similar correlations which need documentation of their basis of equivaleniy with EFPDs.
PROPOSED RESOLUTIONS 1./2./3./4.
PSC should provide necessary clarifications and documentation.
The NRC final draft markup will include suggested changes from Comment 2.
36
NRC COMMENTS - LCO 3.2.2 1.
The second paragraph on page 3/4.2-7, concerning the acceptance of thermocouple readings from Regions 20 and 32 through 37, should be added to SR 4.2.2.
SR 4.2.2 should designate the regions for which COMPARIS0N REGIONS can be used.
2.
The BASIS requires more details and references with regard to the use of COMPARISON REGIONS and experience derived from performing test procedure RT-500K, e.g., specific citation to expanded discussions in Section 3.6 of the FSAR, PSC submittals (some of which appear to be cited in the FSAR), and/or appropriate NRC SERs.
PSC should also address the effect on assumed uncertainties in the regions which rely on a COMPARISON REGION; specifically, in relation to margins in the Hot Spot Evaluation, Section 3.6.4, updated FSAR.
Does the use of COMPARISON REGIONS consume margin in the calculational uncertainties used in the FSAR7 3.
In LCO 3.2.2 ACTION c, the word "immediately" should be replaced with "within 10 minutes".
PROPOSED RESOLUTIONS 1.
The NRC final draft markup will incorporate the suggested revision.
2.
PSC needs to provide the requested references to ensure clarity and to clarify the impact of the use of COMPARISON REGIONS on calculational margin and uncertainty analysis.
3.
The NRC final draft markup will incorporate the suggested change.
NRC COMMENTS - LC0 3.2.3 1.
LCO 3.2.3 would read better if the phrase "for any region used as a COMPARIS0N REGION" were used as the opening rather than the ending of the first sentence of the condition statement.
2.
The BASIS provides no information on the calculational model used to provide calculated RPFs.
Reference should be made to the model description, methods used, and product verification.
3.
A description and references are also appropriate for the technique used in inferring the " measured" RPFs in the COMPARISON REGIONS and their associated uncertainties.
The report GA-A14007, dated February 1977, described a test technique used at low power levels (about 30%)
involving equal flow settings on all orifice valves.
In this case, the temperature rise across the region would be directly proportional to the RPF. At higher power levels, for which equal flow settings on the orifice valves are not feasible, the inference of RPFs would require corrections, based on the orifice valve settings and perhaps on helium circulator speed, number of circulators operating, and/or 37
measured core pressure drop. A reference is needed to quantify the corrections and uncertainties involved in measuring RPFs.
How are uncertainties factored into these determinations? How are any convolutional errors avoided since there are up to seven regions which cannot be used in the evaluation, or is there a normalization required which counts a given COMPARISON REGION more than once?
PROPOSED RESOLUTIONS 1.
The NRC final draft markup will include the suggested change.
2./3.
PSC needs to provide the necessary clarifications and references.
NRC COMMENTS - LCO 3.2.4 This LC0 was deleted from the TSUP (i.e., handled by a separate review action, TAC 52634).
NRC COMMENTS - LCO 3.2.6 1.
LC0 3.2.6 ACTION Statement b requires that "with the maximum POWER-TO-FLOW RATIO greater than 1.17" either reduce power so that the POWER-TO-FLOW RATIO is reduced below 1.17 within 2 minutes plus any delay time defined in ACTION Statement c.1, or "be in SHUTOOWN within the next 5 minutes." 'From LC0 3.2.6 ACTION Statements c.l.b and c.1.c, the additional delay times would appear to be 100 seconds when the POWER-TO-FLOW RATIO is greater than 2.5 and less than 15, and 60 seconds when the POWER-TO-FLOW RATIO is greater than 15.
The BASIS states that the " minimum time to prevent exceeding the SAFETY LIMIT curve of Figure 2.1.1-1 is 2 minutes, which occurs at POWER-TO-FLOW RATIOS of 2.5."
The BASIS further states that "to reach a POWER-TO-FLOW RATIO of this magnitude through an increase in core power, significant equipment malfunction or failure, and/or significant deviations from operating procedurds would have to occur";
the BASIS does not elaborate further.
The BASIS further states that the more liberal time allowances at higher POWER-TO-FLOW RATIOS derive from the fact that the SAFETY LIMIT curve of Figure 2.1.1-1 is based on steady-state analysis of fuel kernel migration, which is really too conservative for the fast transients.
The key issues are that LCO 3.2.6 ACTION Statement b imposes required actions by the operator, and that these actions must be effecte'd in less than 10 minutes.
In the NRC memorandum, B.W. Sheron to R. Fraley (ACRS), date September 13, 1984, the following guidance was stated:
"In order to ensure that the operators are not overburdened, and that sufficient time is available for them to assess the off-nornal condition and take corrective action, we continue to use, as a cuideline, the approach that either:
a) The system should be able to accommodate design basis events without relying on manual action for at least 10 minutes, or b) that automatic features be installed to 38
afford the required protection if the plant cannot acceptably accommodate design basis events assuming no operator action for at least 10 minutes.
I point out that this approach is generally consistent with that recommended in draft ANSI Standard N660 on operator actions times.
(" Time Response Design Criteria for Safety-Related Operator Actions" Draft Standard ANS 50.8/ ANSI N660, Rev. 2, dated March 1981].
"The staff will consider crediting operator action for design basis events in times less than 10 minutes on a case-by-case basis.
For these cases, we ask the licensee or applicant to provide us with applicable data which demonstrates that the actions in question could be reliably taken by an operator within the specified time, and that unacceptable consequences to the health and safety of the public will not occur if the actions are taken at a later (but not excessively later) time than that assumed in the analysis. We are currently using this approach with the Westinghouse Steam Generator Tube Rupture Owner's Subgroup in redefining the times assumed for operator actions during the event.
"In summary, we use the guideline of not allowing operator actions within the first 10 minutes of a design basis event as a means of assuring that the operator will not be overburdened with required actions.innediately after an event occurs, and that features are incorporated into the plant design to assure the operator is not unacceptably burdened.
"When we look at realistic consequences of events, we give full credit to operators taking all expected actions. Moreover, notwithstanding the automatic protective features of the plant, the operator is trained to take actions based on symptoms rather than specific event diagnosis and we expect and encourage operators to react quickly and correctly to events.
For example, if an operator can take an action to activate a needed safety system faster than an automatic system, we encourage the operator to be trained and instructed to do so."
The BASIS for LCO 3.2.6/SR 4.2.6 does not discuss any instrumentation which may be available to the operator to determine the POWER-TO-FLOW RATIO.
LCO 3.3.2.7 and SR 4.3.2.7 describe a nonsafety grade, non-PPS Class I measuring and recording instrumentation system to be used to provide POWER-TO-FLOW RATIO indication by recording onto a paper readout. The BASIS for LC0 3.3.2.7/SR 4.3.2.7 implies that the system is meant to be used primarily for integration of operating times (as required by LC0 3.2.6 ACTION Statement c to ascertain compliance with SL 2.1.1) but not for real time diagnosis of transients.
Per LC0 3.3.2.7 ACTION Statements a and b, the recording system is allowed to be out of service for periods of up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This system is not described in the FSAR. However, the FSAR Section 3.6.7 does describe the core outlet thermocouple scanner, display boards and data i
logger. The scanner is alarmed. Also, on pages 3.6-19 and -20 of the l
update FSAR (Revision 3), power / flow imbalance indication is described as relying on the neutron flux detectors (indication and high flux i
39
scram at 140%; RWP equal to or less than 120% rated power), the rod-out-of-line alarm (2 i 1 ft.), the slack cable alarm, and the steam generator modules reheat steam temperature-high (30*F from loop average) alarm. Nowhere is the information from these document sources integrated to demonstrate that the operator can and will respond to the various POWER-TO-FLOW RATIO indicatinns within less than 10 minutes, and in particular within 2 to 5 minutes.
Violation of the SAFETY LIMIT during power / flow transients is not one of the two DESIGN BASIS ACCIDENTS (events) analyzed for FSV; however, PSC should provide adequate demonstration that violation of the SAFETY LIMIT can be avoided consistent with guidelines quoted above.
The BASIS for LC0 3.2.6/SR 4.2.6 should be clarified accordingly.
As noted previously in NRC Comment 1 to SL 2.1.1, PSC has not provided documentation to support either the integrated or transient curve for the POWER-TO-FLOW RATIO (i.e., Figures 2.1.1-1 and 3.2.6-1 respectively).
Such documentation is needed. However, the staff also notes that none of the Chapter 14 accident analyses addresses power / flow transients in terms of the SAFETY LIMIT POWER-TO-FLOW RATIO.
In particular, the analysis of Helium Circulator Malfunctions (FSAR Section 14.3.2.1) fails to address core response to an anticipated transient without protective action in the form of an automatic turbine-generator runback followed by reactor run back.
This accident could apparently lead to a maximum POWER-TO-FLOW RATIO of 1.54 without operator action, and with no accounting of reactivity feedbacks.
Similarly, Loss of Reactor Generated Steam (FSAR Section 14.3.5.1) without immediate protective action could potentially lead to a high POWER-TO-FLOW RATIO.
Finally, the assumed rod withdrawal transients (FSAR Sections 14.2.2.6 through 14.2.2.8) can result in POWER-TO-FLOW RATI0s of at least 1.4.
However, the accident analysis (GULF-GA-B10872, July 1972) addresses only peak fuel temperatures, not the fuel kernel temperature gradients addressed by the SAFETY LIMITS. PSC needs to update the FSAR to address the SAFETY LIMITS for anticipated transients and other power / flow accidents.
2.
In the BASIS for LC0 3.2.6/SR 4.2.6 (page 3/4.2-26), the statement is made that "the plant control system will usually initiate scram sequences in such cases," 1.e., POWER-TO-FLOW RATIOS above 2.5.
Under what circumstances will the plant control system not initiate scram sequences when the POWER-TO-FLOW RATIOS exceed 2.5? Can a cumulative probability be assigned to these sequences?
l l
PROPOSED RESOLUTIONS l
1.
PSC needs to provide clarification and documentation (in accordance with NRC guidance) on operator response time.
Prior to NRC approval of the final draft, PSC needs to revise the FSAR (Chapter 14) to demonstrate compliance with the SAFETY LIMIT during power / flow transients.
2.
PSC needs to provide requested clarification.
40
NRC COMMENT - LCO 3.3.1 LER 86-009 states, "A revision proposed by the Upgrade Technical Specification Program supports this conclusion by not requiring circulator trip channel operability during similar reactor shutdown conditions." This appears to imply that circulator drain malfunction channel trip OPERABILITY will not be required during SHUTDOWN.
This could not be verified from a review of the final draft Table 3.3.1-3.
PROPOSED RESOLUTION PSC should provide the needed clarification as to the LER statement intention.
NRC COMMENTS - LC0 3.3.2.1 1.
Since operation with the PPS moisture monitors in the " indicate" mode provides no alarm function (it would be provided with the analytical moisture monitors), LC0 3.3.2.1 does not ensure that an alarming condition would be detected in a timely manner.
2.
The BASIS should reference either the FSAR or GA Topical Report whose calculations support the PSC discussion in Attachment 2 to P-85448.
3.
The " alternate monitors" referenced in the BASIS should be identified by equipment number.
PROPOSED RESOLUTIONS 1.
The NRC final draft markup will propose corrections.
2.
PSC should either update the FSAR prior to the amended technical specifications proposal and reference the FSAR in the BASIS, or reference the GA Topical Report.
3.
PSC should incorporate alternate moisture monitor equipment numbers in the BASIS.
PSC should also ensure that the alternate monitors provide an alarm function or provide an appropriate surveillance, as suggested above.
NRC COMMENTS - LCO 3.3.2.2 1.
Table 3.3.2-1, Item 2.b and Table 4.3.2.-1, Item 2.b should be applicable to all modes.
PSC stated (P-85448 Attachment 2) that the monitor is primarily a high-level accident monitor, and its minimum detectable activity would not occur except at power. A monitor is needed for all modes, even if it is reading below its minimum detectable activity.
41
2.
PSC's response in Attachment 2 to P-85448 was incorrect since the statement, "...except surveillances...shall be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after entry into that MODE" was deleted from the April dratt, Specification 4.0.6.
PROPOSED RESOLUTION 1.
NRC final draft markup will make the appropriate changes.
NRC COMMENTS - LC0 3.3.2.3 1.
PSC ACTION Item 23) response is inadequate.
Both the Perry Technical Specification and the STS require 10 days to submit the Special Report.
2.
PSC revised wording in SR 4.3.2.3.2 is contrary to the W-STS intent.
The STS wording is intended to cover all 10 CFR 50, Appendix 8 components / systems, and not just those required to withstand the seismic event (e.g., the ACM, fire protection, and security systems are systems impor tant to safety and should also be included).
3.
PSC did not incorporate the NRC comment in SR 4.3.2.3.2 concerning a CHANNEL FUNCTIONAL TEST. However, Attachment 2 to P-85448 indicated that this has been done.
4.
PSC deletion of seismoscopes is not justified because:
a.
Deletion of previous license requirements is outside the scope of this review.
b.
FSAR, Section 7.3.9, states that the seismoscopes are used for guidance in interpreting the PCRV seismoscope record (i.e., a passive mechanical device used as a backup to an electronic device).
c.
PSC did not commit to the minimum seismic instrumentation required by R.G. 1.12 (i.e., peak accelographs as replacement for I
the seismoscopes).
i PROPOSED RESOLUTION l
1.
PSC should revise their, response to ACTION Item 23).
1./2./3./4.
NRC final draft markup will propose corrections.
NRC COMMENT - LC0 3.3.2.4 PSC response to PSC ACTION Item 25) was unacceptable for the same reasons that their response to ACTION Item 23) was unacceptable (see above).
42
PROPOSED RESOLUTION PSC should revise their response to ACTION Item 25).
The NRC final draft markup will change the 30 days back to 10 days.
NRC COMMENTS - LCO 3.3.2.7 1.
The LC0 3.3.2.7 ACTION Statement for restoration of the permanent recording instrumentation did not agree with that stated in Attachment 2 to P-85448 (i.e., 7 days versus 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).
2.
Maintenance of equipment is not an acceptable basis for ACTION Statements. Also this system is not comparable to post-accident STS monitoring systems since the POWER-TO-FLOW RATIO measurement system is used to verify compliance with an LSSS limit.
3.
Mode changes are appropriate only with the backup recorder in service. As worded, LC0 3.3.2.7, ACTION Statement C. would allow going to POWER with a complete loss of measurement instrumentation within the time frame allowed prior to entering the STARTUP mode.
4.
The FSAR does not describe this system.
PROPOSE 0 RESOLUTION 1.
PSC should revise their response in Attachment 2 to P-85448 to reflect the LCO 3.3.2.7 contents.
The NRC final draft markup will propose corrections.
2.
PSC should revise the BASIS to LC0 3.3.2.7 and address outage times with respect to the probability of an event occurring during the period when redundancy is not available.
The NRC final draft markup l
has delected the nonapplicable information.
i 3.
The NRC final draft markup will propose corrections.
4.
PSC should provide a description of the POWER-TO-FLOW RATIO measurement system in the next FSAR update.
I NRC COMMENTS - LCO 3.3.2.8 j
l 1.
Amendment 28 authorizing the use of comparison regions was reviewed l
based on errors in regions 20 and 32 through 37 due to " jaws" flow I
problems.
There is no NRC-approved justification for the use of this concept for regions with inoperable thermocouples.
2.
FSAR, Section 3.6.7, states that a complete loss of all temperature indication from a single region due to thermocouple failure could be repaired within approximately one half hour.
The LC0 3.3.2.8, i
Action Statement b 24-hour time in which to restore to OPERABLE status, combined with an additional potential 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> between checks in SR 4.2.2, has not been justified.
1 43 l
l l
3.
The SR 4.3.2.8.c requirement states that greater than 2% RATED THERMAL POWER and stable core temperature are needed to perform'the required calibration. Operation at LP or P has not been justified during performance of this calibration.
4.
LCO 3.3.2.8 does not in all cases reflect the wording of FSAR, Section 3.6.7.
PROPOSED RESOLUTION 1./2./3./4.
The NRC final draft markup will propose corrections.
NRC COMMENTS - LC0 3.3.3 Contrary to Attachment 2 to P-85448 and the NRC ACTION Item response provided in the NRC Letter dated October 22, 1985, the action requirements of the BWR-STS were not reflected in LC0 3.3.3.
PROPOSE 0 RESOLUTION i
The NRC final draft markup will propose corrections.
NRC COMMENTS - LCO 3.4.1 1.
LCO 3.4.1 and SR 4.4.1 differ from the updated FSAR (Sections 3.7, 14.8 and 14.11 and Appendix D) in the units and terminology used for
. circulating and plateout activity.
LCO 3.4.1.a and SR 4.4.1.b quote
" primary coolant noble cas beta plus gamma activity" in units of curies-MeV/lb as of '15 minutes after sampling." The activities are normalized to a weighted nuclide beta and gamma average energy, E-BAR.
Iodine circulating and plateout activities are quoted in terms of DOSE EQUIVALENT I-131, which is defined by specification DEFINITION 1.15.
The updated FSAR (Section 3.7 and Tables 3.7-1 and 14.11-3) deals only in units of curies for circulating activity.
In the updated FSAR, circulating activity includes not only noble gases but also Hydrogen-3, halogens, and certain intermediate-mass metallic fission products.
In the FSAR, circulating activities are quoted ct equilibrium values, not 15 minutes decayed, for both the " expected" and " design" fuel release rates.
Table 3.7-1 does provide half-life i
data for each nuclide.
E-BAR is not defined in the FSAR, nor are the nuclide beta and gamma E-BAR values provided in Table 3.7-1.
It j
should be further noted that the circulating activity totals in Tables 3.7-1, and in the first row of Table 14.11-3 differ by up to three percent. Reasons for the difference are not provided in the FSAR.
DOSE EQUIVALENT I-131 is also not defined in the FSAR, and the two methods given in DEFINITION 1.15 for quantifying the DOSE EQUIVALENT I-131 (Thyroid) yield widely disparate results, as noted below:
i 44
1 1
R.G._1.109 Rev 1 (1977) i Nuclide Rads /C1 1-131 Factor mres/oCi I-131 Factor 130 Indeterminate 1.42E-4 0.095 131 1.48E*6 1.0 1.49E-3 1.0 132 5.35E+4 0.036 1.43E-5 0.0096 133 4.00Et5 0.270 2.69E-4 0.181 134 2.50E+4 0.017 3.73E-6 0.0025 135 1.24E*5 0.084 5.60E-7 0.00038
]
2.
SR 4.4.1.a.1 refers to the gross activity monitor, RT-9301, but the 8 ASIS fails to provide an FSAR reference to RT-9301.
The gross activity monitor is described in Section 11.2.3.1.2.1 of the FSAR.
4 I
However, the staf! notes that radiation monitoring instrumentation is i
described in a disjointed and often piece-meal fashion in FSAR i
Sections 7.3.4 and 11.2, both of which lack clear cross-references.
3.
SR 4.4.1.b. refers to " grab samples of primary coolant," but the BASIS fails to describe or reference the applicable sampling and measuring devices and techniques.
Sections 7.3.4 and 11.2 of the FSAR are the i
only available references but provide little or no useful information about the techniques employed.
4.
The licensee's letter P-85443, " Fort St. Vrain fuel Surveillance i
Program," dated November 27, 1985, committed the licensee (as stated on page 4 to Attachment 1) "to perform destructive PIE's (i.e.,
j Post-Irradiation Examinations) on a selected surveillance element following any cycle during which design basis primary coolant activity.
limits as established by Specification LCO 4.2.8 ' Primary Coolant Activity' of the Fort St. Vrain Technical Specifications are reached." The staff notes that the final draft for LCO 3.4.1 (which replaces the current LC0 4.2.8). lacks an ACTION Statement to effect implementation of the cited commitment.
In addition, there is no accompanying SR. The staff also observes that the commitment is so worded in the licensee's letter, P-85443, that there is a strong implication that the destructive PIE will be limited to the most highly exposed elements scheduled for removal from the core at the end of any cycle in which the limits of LCO 3.4.1 may have been reached.
An ACTION Statement derived from this commitment should not be worded to preclude the possibility of a destructive PIE on the most highly thermally or mechanically stressed fuel elements.
Such a possibility should not be precluded even if removal is not scheduled at the end of the affected cycle, or on fuel elements inserted at the beginning of the affected cycle if there is evidence or reason to suspect abnormalities in the manufacturing, shipping, or handling of the fresh elements.
45
5.
SR 4.4.1.c imposes the requirement for a 24-hour grab sample if the primary coolant activity level reaches 25% of the limits of LC0 3.4.1.a. b or c.
The BASIS provides no reasons why SR 4.4.1.c should not apply to the Strontium-90 limit in LC0 3.4.1.d.
In fact, the second paragraph on page 3/4.4-6 of PSC's final draft implies that the "non-measured" Sr-90 level will be calculated or inferred from the grab sample, l
6.
The Basis for Sr-90 and I-131 Activity limits is unclear as to whether the limits are based on the analysis in Section 14.11 of the updated FSAR or on the AEC Safety Evaluation (BASIS Ref. 1).
The BASIS uses the nonstandard terminology "effectivities" in place of " dose factors." The BASIS does not provide specific document citations either for data taken from ICRP II or for data cited for comparison in ICRP VI. However, a comparison of the licensee's inhalation dose factors for I-131 and Sr-90 to those in Table E-7, Reg. Guide 1.109 Revision 1, can be made as follows in units of rems per milli-curie:
BASIS for Reg. Guide 1.109 Nuclide LCO 3.4.1/SR 4.4.1 Thyroid Total I-131 1,480 1,490 1,509 Sr-90 36.700 No Data 14.452 The staff recognizes the evolutionary nature of dose factor utilization in licensing analyses but finds the BASIS and the FSAR unclear as to which set of data is being used for which limits. The degree of conservatism in assumptions needs to be better expressed and more readily determinable.
PROPOSED RESOLUTIONS 1.
PSC should revise the updated FSAR to be consistent with the technical specifications and to provide data and other information needed to interpret the specifications.
In Attachment 2 to P-85448, the argument is made that E-BAR (DEFINITION 1.16) should be limited to the noble gases and not the daughter products.
However, the BASIS states that the limit is based on the entire " design" primary coolant gaseous radioactive inventory (i.e., Table 3.7-1, updated FSAR).
Since this value includes daughter products, E-BAR and the technical specification measurenent should be consistent with the FSAR.
Normalizing factors to yield DOSE EQUIVALENT I-131 should also be defined in both the FSAR and specification DEFINITIONS. Such factors should be used consistently.
2./3.
Specific references to instrumentation should be given in the Bases, and the FSAR should be revised to provide a consistent and detailed description of the coolant, and plateout activity monitors.
1 i
46
O O
4.
PSC should clarify how they propose to track their commitment for fuel surveillance using destructive PIE.
5.
The NRC markup of the final draft will incorporate the proposed corrections.
6.
PSC should provide clarification and appropriate references.
NRC COMMENTS - LC0 3.4.2
)
1.
i The APPLICA8ILITY Statement does not include Operational Modes as defined in Table 1.0-1, and is inconsistent with the format used in LCO 3.4.3.
l 2.
The ACTION Statements are based solely on the prevention of graphite corrosion in the presence of oxidants.
The staff notes that in the sequence of events reported in R0 50-267/84-008, concerning the partial failure to scram on June 23, 1985, high primary coolant moisture levels, or that of other oxidants, may have contributed to the sticking of control rod pairs. Under the increasingly severe 3
conditions of impurity levels covered by ACTION Statements a, b, and c, the demonstration of operability by performing a partial scram test of at least 10 inches on all partially inserted, and fully withdrawn, i
control rod pairs, and of approximately 2 inches on the regulating rod j
pairs, is judged to require an increased frequency over that of at least once per 7 days, as required by SR 4.1.1.c.1 and 2.
See NRC i
COMMENTS - LCO 3.1.1.
3.
SR 4.4.2.1 or other SRs should provide for analysis of the primary coolant initially within one hour following an alarm indicative of the i
presence of primary coolant impurities.
Specifically, alarm actuation on either the CO Infrared Ana.lyzer, the Analytical Moisture Monitoring i
System, the PPS Dewpoint Moisture Monitoring System, or the Primary i
Coolant Pressure-High should be the basis for requiring immediate i
chemical analysis of the primary coolant for ascertaining impurity levels.
]
PROPOSED RESQLUTIONS 1./2.
The NRC final draft markup will propose changes and corrections.
j 3.
PSC should propose appropriate surveillances on alarm of the subject instrumentation.
l 1
NRC COMMENTS - LCO 3.4.3 1.
There is no explanation either in the BASIS or in the updated FSAR as to the method for correlating the moisture dewpoint limits with ppm by volume. Further, there is no statement defining the equivalency with 1
j the limits of LCO 3.4.2 at 1200*F.
Presumably a dewpoint of -20*F i
I I
47 i
,--.--n-----,---.-.--------,.m------,,--,------..----.a--e-,.
..------g.w,-en
-,,-w.,
y ---,,,- -,,,,- -, - -
corresponds to 10 ppm (by volume) at 1200*F, but this is inferred from, rather than explicitly stated in, figure 3.4.3-1.
Similarly, there are abrupt jumps from 10 ppm to 1,000 ppm (by volume) for CO2 and from 10 ppm to 15,000 ppm (by volume) for C0 at 1200*F between LCO 3.4.2 and LC0 3.4.3.
Clarification is needed to ensure that these large changes are both appropriate and conservative.
SHUT 00WN should also be included under APPLICA8ILITY.
2.
The BASIS should include FSAR references for all the limits.
In particular, the technical basis for boron carbide oxidation rates is not found in the updated FSAR.
3.
The ACTION Statements require correlation with control rod operability limits of LCO 3.1.1.
4.
SR 4.4.3 or other SRs should provide for analysis of the primary coolant initially within one hour following an alarm indicative of the presence of primary coolant impurities. At a minimum, the response to alarm actuations should include those of the C0 Infrared Analyzer, the Analytical Motsture Monitoring System, the PPS Dewpoint Moisture Monitoring System, and the Primary Coolant Pressure-High.
PROPOSED RESOLUTIONS 1./2./3.
PSC should provide the additional information and needed clarification and_ changes.
4.
PSC should propose appropriate surveillances on the alarm instrumentation.
NRC COMMENTS - PRIMARY COOLANT (OTHER) 1.
Both the Helium Purification Dryer and the Low Temperature Adsorber (LTA) in the Helium Purification Train must be available for a period of five hours during the Design Basis Accident No. 1 (Permanent Loss of Forced Cooling) to effect depressurization and cleanup of the
(
released primary coolant gases.
The assumed availability of these specific components is specified in the BASIS for LC0 3.7.5 (page 1
3/4.7-27) and described in Section 14.10 and Appendix 0 of the updated FSAR. However, there are no LCO's which correlate primary coolant
' impurity levels with the remaining life in the Helium Dryer before required shutdown for regeneration.
Since the plant has experienced several occasions in which the Helium Purification Dryer and Low Temperature Adsorber in each loop were successively incapacitated by high moisture levels, there is a need to predict the remaining lite in the Helium Purification Dryers so that shutdown can be effected with a minimum of five hours in lifetime remaining prior to requiring regeneration.
l 48 l
-,m
-,,. -. + -, - -
--,,.--.y-.wr----,n-.-m-,--,--w-w-
m 2.
Although the reactor has recently experienced component failures from chloride-induced stress corrosion cracking of steel component 1
materials, there are no LCO's on allowable chloride impurity levels in the primary coolant; nor are there SR's for tracking chloride-induced corrosion. The replacement of certain steel components with Inconel materials has been made by PSC to " reduce" the effects of chloride-induced corrosion; however, there are numerous metallic reactor components, including the carbon steel liner, which may require periodic surveillance to ensure that chloride-induced l
corrosion is either not occurring or occurring at a rate which can be monitored to prevent unexpected damage to reactor components.
1 3.
The presence of moisture and condensed water in the purified helium purge and pressurization flow lines has possibly been a major contributor to failures ir the CRDMs and RSD units.
Since purge flow enters the primary coolant through the CRD cable penetration! this i
flow can be a source of exidants with potentially high local levels of concentration before dispersal within the large primary coolant volume. The incorporation of NRC Comments 2 and 3 to LCO 3.1.1 will provide one line of defense against moisture ingress into the primary coolant from CRD purge flow.
However, there are currently non-safety moisture sensors in the lead lines to, and drain lines from, the i
helium circulator and steam generator penetration interspaces where purge flow moisture contamination is most likely to occur (based on documented experience). As indicated in greater detail in NRC Comment 2 to LCO 3.6.1.3, there is ample reason to believe that the helium in the PCRV penetration interspaces should.be subject to moisture impurity limits (such a limit is already imposed by Table 7.1-1 of the FSAR, but not enforced), and that these limits should require appropriate LCOs for the moisture-sensing devices referred to above.
The incorporation of additional proposed LCOs will provide two lines of defense against moisture ingress into the primary coolant from purge flow.
PROPOSED RESOLUTIONS l
1./2./3.
PSC should provide the reasons why such LCOs are not needed.
NRC COMMENTS - LCO 3.5.4 1.
PSC deleted first draf t LC0 3.5.4.b without justification.
2.
As worded, the final draft LC0 3.5.4.d does not ensure an OPERABLE flow path to the instrument air compressor and emergency diesel generator coolers.
3.
SR 4.5.4.f.2 of the first dr~ aft was deleted from the final draft without justification.
4.
It is unclear as to how the SR 4.5.4.1.e system flush is performed.
49
5.
In accordance with the PSC letter P-85488, dated December 20, 1985, 1500 gpm is a fire protection safe shutdown basis.
PROPOSED RESOLUTION l./2./3./5.
NRC final draft markup will propose corrections.
4.
LCO SR 4.5.4.1.e should be revised to provide clarification.
NRC COMMENTS - LC0 3.6.1.1 1.
PCRV safety valve settings are given in psig as referenced and listed in Table 2.2.1-1.
LCO 3.6.1.1 is applicable "whenever PCRV pressure exceeds 100 psia" which implies that pressure indication is in psia.
Because of the usage of different units, are there any circumstances in which either this LCO or other implementing procedures could cause confusion to plant personnel, thereby leading to operation in a manner less conservative than specified by either SL 2.2.1 or LC0 3.6.1.17 2.
NRC had previously marked the April 1985 draft of SR 4.6.1.1.a. b, and d to include CHANNEL FUNCTIONAL TESTS and CHANNEL CALIBRATIONS.
Those changes are not reflected in the final draft and are not explained as exceptions in Attachment 2 to P-85448.
3.
On page 3/4.6-3 of the BASIS, the statement is made that " verification that the interspace pressure is less than 5 psig may be made by absence of alarm or other means." As long as " absence of alarms" is indicative of low pressure and not an alarm failure, this statement is 3
true, but "other means" needs further explanation, t
4.
The ACTION Statement permits 12 hrs for restoration and then 24 hrs to l
SHUTDOWN versus 15 minutes for restoration, and 6 hrs to SHUTDOWN in the STS Section 3.4.2.2 ACTION (closest equivalent to PCRV safety valves and penetration overpressure protection).
PROPOSED RESOLUTIONS 1.
PSC needs to provide clarification.
2.
PSC needs to include the CHANNEL FUNCTIONAL TESTS and CHANNEL CALIBRATIONS as indicated in the NRC markup of the April 1985 draft to be consistent with the STS.
l 3.
PSC needs to provide clarification.
4.
PSC should change their ACTION times to agree with the STS, or justify why not.
t
]
50
NRC COMMENTS - LCO_3.6.1.2 1.
PSC ACTION Item 34) response addressed FSAR, Section 6.8.1, which is for PCRV safety valves, instead of the LCO 3.6.1.2 steam generator and i
circulator penetration relief valves. Therefore, the response is unacceptable.
2.
The LCO 3.6.1.2 BASIS should clarify "other means" in the first paragraph on page 3/4.6-7.
3.
A new LCO SR 4.6.1.2.b was added to verify each 31 days that the block valve is open.
This apparently addresses the NRC's first draft comment on block valve position verification, as documented in the NRC markup of the April 1985 draft of SR 4.6.1.2.e.
However, Attachment 2 to P-85448 did not identify this as being the case.
4.
In the April 1985 draft of SR 4.6.1.2, NRC marked in CHANNEL FUNCTIONAL TESTS.ind CHANNEL CALIBRATIONS.
These are not reflected in the current draft.
5.
The operability requirements for the steam generator and circulator penetration overpressure protection trains need to be expanded to include:
a.
Verification that the block valve in the OPERABLE train is open after maintenance or testing.
b.
Appropriate instrumentation surveillances.
6.
In the BASIS, the final sentence of the first paragraph is out of context.
7.
The ACTION Statement permits 12 hrs for restoration, and then 24 hrs to SHUTDOWN versus 15 minutes for restoration, and 6 hrs to SHUTDOWN in the STS Section 3.4.2.2 ACTION (closest equivalent to PCRV safety valves and penetration overpressure protection).
PROPOSED RESOLUTIONS 1.
PSC should revise their response to ACTION Item 34).
2.
The LC0 3.6.1.2 BASIS should be revised.
3.
PSC's Attachment 2 to P-85448 should be revised.
4.
The NRC final draf t markup will make the required changes.
5./6.
The NRC final draft markup will propose corrections.
7.
PSC should change their ACTION times to agree with the STS, or justify i
l why not.
l l
51
a t
1 NRC COMMENTS - LCO 3.6.1.3 1.
In the markup of the April 1985 draft, for SR 4.6.1.3, NRC had indicated the need for CHANNEL FUNCTIONAL TESTS and CHANNEL CALIBRATIONS. The SR has been rewritten to delete the affected surveillances, but there is no justification for these exceptions in i
l to P-85448.
l 2.
LCO 3.6.1.3, which applies only when "PCRV pressure exceeds 100 psia,"
addresses the needs to assure prevention of "any leakage of contaminated helium from the primary coolant system through the primary closure" and to " ensure that any leakage through the secondary l
closure into the reactor building will be purified helium."
In the event of primary closure leakage, additional radiological protection is provided by compliance with LCO 3.6.1.4 and LC0 3.6.1.5.
As noted in the BASIS for LC0 3.6.1.2/SR 4.6.1.2, "the steam generator and circulator penetrations are the only PCRV penetrations that contain process fluids" at high pressure. The process fluids include feedwater, high pressure steam, reheat steam, and circulator bearing water. LCO 3.6.1.3 and SR 4.6.1.3 provide a mechanism for preventing reheat steam inleakage into the steam generator penetration interspaces through identified leak paths, at least, when PCRV pressure is above 100 psia.
The prevention of the inleakage of other process fluids is not dependent on LC0 3.6.1.3, but rather upon the sign of the pressure differential and on whether the leak rate is sufficient to actuate a loop shutdown or a circulator trip on high I
pressure in the penetration interspaces (LCO 3.3.1, Tables 3.3.1-2 and 3.3.1-3).
At PCRV pressures of less than 100 psia, LC0 3.6.1.3 does not apply, and so will not prevent the inlea.kage of steam originating in either the flash tanks or auxiliary boilers, and routed to the circulator turbine drives through part of the reheat steam modules.
Under the same conditions (to which again LCO 3.6.1.3 does not apply) 4 there appears to be no mechanism to control or preclude inleakage from circulator bearing water pipe leaks. Moisture and water in the penetration interspaces is a possible source of water which can be i
transported throughout the PCRV purified helium purge and pressurization system. This source is suspected to have contributed to control rod drive sticking (R0 50-267/84-008-03), to material i
degradation within the RSS (R0 50-267/84-012-03), and to corrosion and blockage in other PCRV helium purge and pressurization lines (R0 50-267/85-007).
The NRC-proposed revisions to LC0 3.1.1 and LC0 3.1.6 should ensure that these events are not repeated. However, the presence of water in the purified helium purge and pressurization system violates the assumption of dryness (zero humidity) in the PCRV Instrument Penetration Interspaces, as defined in Table 7.1-1 of the updated FSAR (Revision 3). Also, although LC0 3.6.1.4 provides a means to establish the acceptability of potential leak rates through l
primary and secondary closures, the presence of moisture in the l
penetration interspaces may contribute to the corrosion of closure component materials which are not subject to inspection.
Thus, the l
52 l
a s
lack of dryness in the penetration interspaces may be a degradation of I
defense-in-depth, since means are available to detect moisture ingress and respond accordingly. These are apparently not being used effectively as noted below.
There has been a succession of other events which have demonstrated 1
PSC's inability to respond to the available non-safety indication of j
moisture ingre:s into the PCRV penetration interspaces. As illustrated by RO 50-267/81-067, R0 50-267/85-007, R0 50-267/85-013, and PSC's letter P-84418, dated October 29, 1984, FSV's operators do not appear to respond either to the alarmed moisture probes in the helium circulator and steam generator penetration interspaces or to the more recently installed level indications and alarms available from knockout pots on the drain lines from the respective penetration interspaces. RO 50-267/84-012-03 cites the latter instrumentation as one of several corrective actions.
However, R0 50-267/85-013 indicates that, as long as these alarmed indications are not subject to an LCO, i.e., as long as they are non-safety rated, they may be ineffective as a first line of defense-in-depth.
3.
ACTION permits 24 hrs for restoration, and then 24 hrs to SHUT 00WN.
The closest STS equivalent (Section 3.4.2.2) is much more restrictive.
PROPOSED RESOLUTIONS 1
1.
The NRC final draft markup will reinsert the required surveillance.
2.
PSC should consider additional LCOs to address this continuing problem.
l 3.
NRC final draft markup will change 24 hrs to 12 hrs for SHUTDOWN.
NRC COMMENTS - LCO 3.6.1.4 l
1.
In the markup of the April 1985 draft of SR 4.6.1.4, NRC had requested i
the inclusion of CHANNEL FUNCTIONAL TESTS and CHANNEL CALI8 RATIONS.
In Attachment 2 to P-85448, PSC cites consistency with W-STS, but this position is incorrect since W-STS does require surveillances for the specific instrumentation used to monitor limits in the LCO.
2.
The BASIS makes reference to leakage and leak rates through both primary and secondary closures. The FSAR discusses preconstruction verification tests of leak rates through PCRV liner attachment welds (Section 5.7.2.3) and PCRV closure seals and gaskets (Section 2
5.8.2.3).
In the case of the liner attachment welds, test data are given.
In the case of the seals and gaskets, the test data are stated to be given in the FSAR but cannot be located in either Sections 5.8.2.3, 5.12 or 5.13.
Only the test specification limits for seals and gaskets are used to estimate total expected leakage rates through secondary closures.
Does this mean that the test data were at the j
test specification limit or possibly exceeded it7 53 J
In both Sections 5.7.2.3 and 5.8.2.3, values of leak rates at test conditions (about 2-5 psid) were extrapolated linearly with differential pressure to plant operating conditions (about 688 psid).
In neither case, did the FSAR explain that the laminar flow assumption was being employed.
In both cases, however, the volumetric flow was properly extrapolated based on the laminar flow assumption.
- However, the mass flows are misquoted because the conversion from volumetric flow used a helium density (about 0.01 lbm/cu ft.) at atmospheric pressure rather than the operating densities in either the penetration interspaces (about 0.45 lbm/cu ft.) or the reactor (about 0.23 lbm/cu ft.).
The discussion of exceptions to NRC suggestions as given in to P-85448 makes use of such incorrect mass leak rates in reference to expected secondary closure leakage.
3.
The FSAR cites PCRV leakage of 100% of the reactor primary system volume per year as being used for " des.ign purposes" (FSAR Section 5.2.3) and " design basis" (FSAR Appendix D.1), but the 100% per year value yields leak rates less than the limit of LCO 3.6.1.4.
The higher leak rate in LC0 3.6.1.4 is indicated to account and allow for secondary seal degradation (FSAR Section 5.12.2).
The " design" leak rate of 100% per year is implied to have a nonradiological significance, specifically helium inventory accounting purposes.
However, the " design basis" 100% per year leak rate is used for comparison in Appendix 0.1 to justify conservatisms assumed in the analysis of DBA No. 1.
FSAR Section 5.13.2.3 states that the
" allowable leak rate from the PCRV" is based on "the design primary coolant radioactivity" and is limited to.14.4% by volume per year to meet the requirements 10 CFR 20.
PSC is inconsistent in the usage of the word " design" to refer to nonradiological consequences with regard to leak rates, while ref. erring to radiological limits in terms of coolant inventory.
PSC should resolve such inconsistencies in word usage in the BASIS and FSAR. Also, is the 14.4% per year " allowable limit" still consistent with the current version of 10 CFR 207 4.
ACTION permits 24 hrs for restoration, and then 24 hrs to SHUTDOWN.
r The closest STS equivalent (Section 3.4.2.2) is much more restrictive.
i PROPOSED RESOLUTIONS l
l 1.
NRC final draft markup will add back the required tests.
2./3.
PSC needs to provide clarification and corrections to the BASIS and the FSAR.
4.
NRC final draft markup will change 24 hrs to 12 hrs for SHUTDOWN.
NRC COMMENTS - LC0 3.6.1.5 1.
Item 3.6.1.5.b and ACTION Statement b.
200 CPM is not a definite measurement of gross activity.
54
2.
ACTION Statement a.1.
PSC changed the sampling requirement without justification.
3.
ACTION Statement a.3 was added by PSC without justification.
ACTION Statement a.3. appears redundant with the APPLICASILITY Statement.
PSC can always choose to increase steam generator penetration interspace pressure above primary coolant pressure.
4.
ACTION Statement b.
The last sentence beginning "If Leakage Exceeds" is redundant as the ACTION of specification 3.6.1.4 is just a repetition of the ACTION just called out in b, of 3.6.1.5.
PROPOSED RESOLUTION 1.
PSC should specify gross activity in units of microcuries per milliter.
2.
NRC final draft markup will retain the wording of the April 85 draft for a.1.
3.
NRC final draft markup will delete ACTION Statement a.3.
4.
NRC final draft markup will delete the last sentence in ACTION Statement b.
NRC COMMENT - LCO 3.6.4
- 1. to P-85448, ACTION Item 1 response, is not adequate justification.
Reference is made to the W-STS LC0 3.6.1.7 which is just for structural integrity.
PSC's LCO 3.6.4 also covers primary pressure boundary integrity. A separate action should exist for each component of this LCO (i.e., one for structural and one for pressure boundary integrity).
The requirements of W-STS, LCO 3.6.1.1 are applicable.
2.
Specifications 4.6.4.3.d and 4.6.4.4.b did not incorporate previous NRC comments. Attachment 2 to P-85448 also did not provide justification for not incorporating the NRC comments.
'l t
3.
Specification 4.6.4.1.d did not incorporate the six month reporting frequency required in the staff's SER dated July 8, 1985, and PSC's letter P-85199, dated June 14, 1985.
4.
Specification 4.6.4.1.1.a.1 did not reflect the commitments outlined in the licensee's letter P-85199.
5.
Specification 4.6.4.1.a.2 did not reflect the commitments outlined in the licensee's letter P-85071, dated March 5, 1985.
6.
The Table 4.6.4-1 definition for "new tendons" did not reflect the wording in P-85071.
7.
Tables 4.6.4-1 and 4.6.4-2 did not define the inaccessible tendons.
55
8.
Page 3/4 6-28 definition of " failed wire" did not totally agree with that submitted in P-85071.
)
9.
The acceptance and reporting criteria on Page 3/4 6-28 were not in accordance with that previously accepted in the staff's July 8, 1985, SER, as outlined in the licensee's letter P-85071.
PROPOSED RESOLUTIONS 1.
In addition to the NRC final draft markup proposed corrections, PSC should revise LCO 3.6.4 to ensure:
i a.
Actions reflect both structural and pressure boundary integrity, b.
Specifications for tendons (4.6.4.1), load cells (4.6.4.2), PCRV concrete (4.6.4.3), PCRV Liner (4.6.4.4), and penetrations (4.6.4.5) contain acceptance criteria.
The surveillance should contain a timeframe in which engineering c.
l evaluations are performed once a degraded condition is identified.
2.thru 9.
The NRC final draft markup will propose corrections. PSC should i
provide the appropriate numerical value as required by NRC.
s NRC CONMENTS - LCO 3.6.5.1 i
1.
ACTION time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is less restrictive than the STS time to be in l
SHUTDOWN of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and of the April 1985 draft of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
PSC 1
did,not present any justification for this change.
2.
ACTION under SHUTDOWN and REFUELING,"
.. control rod movements resulting in positive reactivity changes..." is by itself too specific.
There are other changes, such as temperature reduction, which also result in positive reactivity. Also the last "and" was changed to "or" without justification.
3.
Surveillance 4.6.5.1.a.1 on reactor building overpressure has had words
" except as required by specification 4.6.5.3" deleted out of the April 85 draft without explanation.
4.
Surveillance 4.6.5.1 deletes the requirement of the existing FSV Technical Specification (4.5.1.a)5.) without explanation.
5.
The BASIS does not address the surveillance item on system louvers.
6.
Reactor building confinement integrity should be defined as in the W-STS and existing LCO 4.5.1.
i j
1 k
I i
56
O 4
PROPOSED RESOLUTION '
(
l 1.
In the ACTION, the NRC final draft markup will change 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> back to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
In the ACTION under SHUTOOWN and REFUELING, the NRC final draft markup will change " control rod movements" to "any evolutions" and the last "or" to "and".
3.
NRC final draft markup will add the words "Except as required by specification 4.6.5.3.a" 4.
PSC should add back existing FSV Tech. Spec. Item 4.5.1.a).5 and update the BASIS appropriately.
5.
PSC should add a BASIS Statement for the system louvers.
6.
The NRC final draft markup will propose corrections.
NRC COMMENTS - LCO 3.6.5.2 1.
ACTION a and b under SHUTDOWN and REFUELING,"... control rod movements resulting in positive reactivity changes..." is by itself too specific.
There are other chaages, such as temperature reduction, which also result in positive reactivity.
2.
ACTION under SHUTOOWN and REFUELING required immediate suspension in the April 1985 draft versus suspension after 7 days in the November 1985 draft. PSC did not justify this change.
3.
In the APPLICA8ILITY and ACTION, the asterisk on SHUTOOWN does not appear to be justified.
4.
BASIS p 3/4 6-48. PSC bases for the ACTIONS and SURVEILLANCES per the NRC request on the April 1985 markup are not adequate since specific bases are not given.
5.
Specification 4.6.5.2 does not appear to be in agreement with the STS concerning the methyl iodide penetration calculation.
6.
It is unclear, from a review of the original amendment authorizing the 4400-hour Specification 4.6.5.2.c requirement, as to the justification for deviation from the Regulatory Guide 1.52 requirement of 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />.
7.
The LCO and ACTIONS should address OPERABLE trains instead of components.
8.
Specification 4.6.5.2.b of the first draft was deleted without justification.
9.
Specification 4.6.5.2.b added additional words concerning painting with low solvent paints without justification.
57 i
- 10. The first draft bypass leakage value of Specification was revised without justification.
- 11. The first draft flow rate acceptance criteria was revised in Specification 4.6.5.2 without justification.
12.
Subatmospheric should be quantified in the BASIS.
PROPOSED RESOLUTION 1.
NRC final draft surkup will change " Control Rod Movements" to "Any Evolutions."
2.
PSC should justify the change from immediate suspension to 7 days, or leave the action as immediate suspension (10 minutes).
3.
NRC final draft markup will delete the asterisk on SHUTDOWN.
4.
PSC should provide specific bases for the ACTIONS and SURVEILLANCES.
5.
The licensee should clarify what formula and data were used to calculate methyl lodide penetration.
6.
PSC should provide the justification for this deviation.
If the 4400-hour frequency is based on operational experience, PSC :hould provide sample surveillance results and the acceptance criteria.
7./8./9./11./12 The NRC final draft markup will propose corrections.
10.
PSC should provide the required justifications.
i NRC COMMENTS - LCO 3.6.5.3 1.
Existing FSV Tech, spec. prerequisites for quarterly louver testing l
(SR 5.5.2, p. 5.5-2) are omitted from the November 1985 draft without justification.
l 2.
Surveillance 4.6.5.3.b.1 changed "...each louver group opens fully..." in the April 1985 draft to "...each louver group opens as i
designed..." in the November 1985 draft.
FSAR Section 6.2.3.4
- p. 6.2-12, states that the louvers are fully open within approximately I second.
PROPOSED RESOLUTION 1.
PSC should add the existing FSV Tech. Spec. prerequisites on quarterly i
louver testing or justify the deletion based on incorporation of the information into surveillance procedures (outside the Tech Specs).
2.
NRC final draft markup will change"...each louver group opens as designed..." to "...each louver group opens fully...".
l l
58 L
- 10. The first draft bypass leakage value of Specification was revised without justification, i
- 11. The first draft flow rate acceptance criteria was revised in l
Specification 4.6.5.2 without justification.
- 12. Subatmospheric should be quantified in the BASIS.
PROPOSED RESOLUTION 1.
NRC final draft markup will change " Control Rod Movements" to "Any Evolutions."
2.
PSC should justify the change from immediate suspension to 7 days, or leave the action as immediate suspension (10 minutes).
3.
NRC final draft markup will delete the asterisk on SHUTDOWN.
4.
PSC should provide specific bases for the ACTIONS and SURVEILLANCES.
I 5.
The licensee should clarify what formula and data were used to calculate methyl todide penetration.
6.
PSC should provide the justification for this deviation.
If the 4400-hour frequency is based on operational experience, PSC should provide sample surveillance results and the acceptance criteria.
7./8./9./11./12 The NRC final draft markup.will propose corrections.
- 10. PSC should provide the required justifications.
NRC COMMENTS - LCO 3.6.5.3 4
i 1.
Existing FSV Tech. spec. prerequisites for quarterly louver testing (SR 5.5.2, p. 5.5-2) are omitted from the November 1985 draft without i
justification.
2.
Surveillance 4.6.5.3.b.1 changed "...each louver group opens fully..." in the April 1985 draft to "...each louver group opens as l
designed..." in the November 1985 draft.
FSAR Section 6.2.3.4,
- p. 6.2-12, states that the louvers are fully open within approximately 1 second.
i i
PROPOSED RESOLUTION 1.
PSC should add the existing FSV Tech. Spec. prerequisites on quarterly louver testing or justify the deletion based on incorporation of the information into surveillance procedures (outside the Tech Specs).
l 2.
NRC final draft markup will change"...each louver group opens as designed..." to "...each louver group opens fully...".
i 58 1
O kRC COMMENTS - LCO 3.7.1.1 1.
ACTION Statement on restoring OPERABILITY within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> does not recognize condensate or boosted firewater which may be used as backup to the boiler feed pumps.
2.
In ACTION Statement b. "or control rod movements..." is too specific.
3.
PSC has not explained why an operating auxiliary boiler is not required in part (a) of this LC0" (see E.J. Butcher LTR to 0.R. Lee dated October 22, 1985, NRC 10).
PROPOSED RESOLUTION l
1.
PSC should consider verifying the operability of the condensate and/or boosted firewater during the 72-hour restoration period.
2.
NRC final draft markup will change "...or control rod movements..."
to "...and any evolutions...".
3.
PSC should expla'in why an operating auxiliary boiler is not required in part (a) of this LCO.
NRC COMMENTS - LCO 3.7.1.2 1.
Item 3.7.1.2.c dump tank safety valves setpoint has been changed from 860 1 10 psig in the April 1985 draft to " equal to or less than 870 psig" in the November 1985 draft. Without a lower-limit setpoint, l
venting and draining of the tank to the radioactive gaseous and liquid systems could occur before the contents have been adequately cooled, as stated in the BASIS.
2.
STARTUP was omitted from the APPLICABILITY per PSC justification in P-85448 Attachment 2.
PSC states that the average graphite temperature in STARTUP is less than 500*F, well below the 900*F threshold for significant steam graphite reaction (FSAR Section 14.5 and Appendix A.12).
PSC has not indicated what graphite temperatures in STARTUP might be under accident conditions.
3.
In the ACTION, PSC changed the restoration time from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days, and STARTUP was not changed to SHUTDOWN per the NRC April 1985 markup.
In the BASIS, PSC states that operating experience and the small probability for the design accident justify a 7-day restoration time which may be required for repairs.
The 11 years operating experience may be atypical because of infrequent power operation.
The small probability of the design accident and estimated repair time are not adequate justification for a restoration time.
59
4.
SR 4.7.1.2.c.2.
Contrary to the NRC April 1985 draft markup, PSC has changed " block" valve to " isolation" valve and has omitted "on receiving actuation signal".
5.)
Existing FSV Tech. Spec. surveillances on dump tank pressure and temperature on p 5.3-1 have been omitted without fjetificaticn.
6'.
SR 4.7.1.2.c. is not adequately discussed in the BASIS.
Restating the surveillance requirement in the BASIS does not constitute a basis for the surveillance.
7.
Contrary to'PSC's response for not incorporating NRC comments in Attachment.2 to P-85448, the steam / water dump tank activity monitors (shi.ca ;ihitors RIS-93250-12 and RIS-93251-12) are required in Specification 4.3.2.2.
The licensee's respor.se stated that OPERABLE dump tank activity monitors did not have to be specified.
8.
If a parameter (e.g., water level) is called out in the LCO, then an appropriate surveillance is required.
9.
The action statement, as worded, does not appear to be appropriate for a steam / water dump tar.t water level greater than 45 inches, when compared to the existing FSV LC0 4.3.3 or W-STS.
PROPOSED RESOLUTIONS 1.
NRC final draft markup will change the setpoint back to 860 i 10 psig.
2.
PSC should justify why an accident in STARTUP would not lead to graphite temperatures greater than 900*F and why the steam / water dump system is not needed in STARTUP during power reduction from POWER to LOW POWER to STARTUP when the graphite temperatures might still be high.
3.
NRC final draft markup will change restoration time to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; STARTUP will be changed to SHOTDOWN; the BASIS will be changed accordingly.
4.
NRC final draft markup will make the appropriate changes.
5.
PSC should justify the deletions.
6.
PSC should rewrite the BASIS to justify the specific surveillance, not just restate it.
7.
PSC should revise their response.
8./9.
The NRC final draf t snarkup will propose corrections.
60
NRC COMMENTS - LCO 3.7.1.3 1.
The BASIS does not agree with the FSAR, Revision 3, Section 10.2.5.3, in that the HRH PORV now actuates on loss of degraded voltage instead of on a turbine trip.
2.
It is unclear as to the purpose of Specification 5.2.3 of the final draft.
This section does not list all the steam safeties [e.g., main steam electromatic safety valves (one/ loop), superheater safety valves (three/ steam generator), reheater safety valves (one/ steam generator),
deaerator safety valves (two), etc...),
PROPOSED RESOLUTION 1.
The NRC final draft markup will propose corrections.
2.
PSC should provide clarification.
NRC COMMENTS - LCO 3.7.1.4 As in the STS, the intent is to be in a mode that reduces the possibility of an inadvertent radioactive atmospheric release due to high secondary activity.
Therefore, unless the licensee's philosophy of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to STARTUP and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to SHUTDOWN is defined in the technical specifications, the requirements, as worded, would authorize operation at POWER for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scram.
PROPOSED RESOLUTIONS The NRC final draft markup will propose corrections.
NRC COMMENTS - LCO 3.7.2 1.
ACTION Statement a.
PSC changed the proposed April 1985 markup to
"... Isolate the affected secondary coolant within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />," from "...be in SHUTOOWN and isolate the affected secondary coolant loop within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />." The existing FSV Tech. Spec. LC0 4.3.7. also requires SHUTDOWN either immediately or within one hour.
2.
PSC changed from the April 1985 markup of "12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />" to "24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" without justification.
3.
CHANNEL CALIBRATION is done once per 18 month but the existing FSV Tech. Spec. SR 5.3.5 requires once per year.
61
4.
BASIS. Third paragraph discusses reactor SHUTDOWN contingent upon hydraulic fluid pump and/or accumulator restoration. Howaver, the ACTION only discusses valve accumulator loss, or loss of 2500 psig.
Since there is a specification on hydraulic pump OPERABILITY (3.7.2.c.) and one on maintaining 2500 psig (3.7.2.b), PSCs discussion of the ACTION in the BASIS is not clear.
There apparently is no ACTION connected with loss of hydraulic pumps (3.7.2.c), unless that leads to loss of 2500 psig (3.7.2.b).
5.
It is unclear as to whether the ISO-degree F limit is to ensure system operation or for aging considerations.
6.
It is unclear why:
(1) the FSAR and existing Technical Specifications are in error concerning existing Technical Specification Actions for isolating the non-affected loop versus the affected loop; (2) the Reference Design Manual states that the hydraulic pump (e.g.,
emergency pump) is required to keep the loop shutdown; and (3) the FSAR, Revision 1, Section 9.11.3-2, states that the emergency hydraulic pump is needed to keep the valves in position during a loop shutdown.
This proposed change also does not agree with Amendment 10, dated January 27, 1976, which reversed the original procedure from a shutdown of the affected loop to shutdown of the non-affected loop.
7.
It is unclear:
(1) what fast-acting secondary coolant system hydraulic valves were assumed to be operable as an initial condition in the FSV accident analysis; (2) in what LC0 their OPERABILITY is explicitly identified; and (3) whether or not stroking time was an initial condition requirement for the FSV accident analysis.
8.
It is unclear as to the definition of "immediate threat," as referenced in Attachment 2 to P-85448.
9.
It is unclear how the affected loop is isolated upon a loss of one system.
10.
Specification 3.7.2.c does not appear to have a corresponding ACTION, therefore, with less than two OPERABLE hydraulic pumps, Specification 3.0.3 would apply.
- 11. The 150-degree F LCO requirement should have a carresponding surveillance.
PROPOSED RESOLUTION 1.
NRC final draft markup will change to cause SUUTDOWN and isolation within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
PSC should rewrite the BASIS for the ACTION accordingly.
2.
NRC final draft markup will change appropriately.
3.
PSC should justify why CHANNEL CALIBRATION was changed from one year to 18 months on SURVEILLANCE.
62
4.
PSC should clarify the BASIS to correspond to the ACTION.
5.thru 9.
PSC should provide the required clarifications to the above comments.
- 11. The NRC final draft will propose corrections.
NRC COMMENTS - LCO 3.7.3 1.
The existing FSV Tech. Spec. SR 5.3.6 or, monthly FUNCTIONAL TESTS and annual calibrations of pressure indicators and low-pressure alarms on the instrument air receiver tanks and headers was deleted without justification.
2.
The BASIS discussion of the ACTION and SURVEILLANCE requirement is nonspecific.
Such BASIS discussion could be used almost word for word for most ACTIONS or SURVEILLANCES.
Such nonspecific discussion defeats the purpose of a BASIS discussion.
Specific references to FSAR Sections 9.9 and 10.3.9, as in the existing FSV BASIS discussion, should also be considered.
PROPOSED RESOLUT70N 1.
NRC final draft markup will add back the surveillance.
2.
PSC should write the BASIS for the ACTION and SURVEILLANCE requirements so that explicit mention is made of why or how the requirements meet the performance described in the safety analysis.
NRC COMMENTS - LC0 3.7.4.1/3.7.4.2 1.
FSAR, Section 9.7.4.1, Page 9.7-12 and the Design Reference Manual SD-46, Page 11, confirm that firewater is not a backup to the PCRV cooling water heat exchangers, contrary to LC0 3.7.4.1, Action c.2 and the BASIS.
2.
The circulating-water tower basin can serve as a backup for all service water loads via a cross-connect between the service water and circulating-water systems.
However, LCOs 3.7.4.1 and 3.7.4.2 do not address this capability.
LCO 3.7.4.1.c covers the path from the circulating-water pumps through HV-4153 to the service-water pump pit. LCO 3.5.4 covers the path from the circulating-water makeup pumps to the firewater pump pit. However, the path from the circulating-water tower basin through HV-4 221-1/ -2 to the service-water pump pit is not addressed.
l 3.
LCOs 3.7.4.1 and 3.7.4.2 ACTION do not indicate that inoperable components serviced by the service-water system affect LCOs 3.6.2.1, l
3.6.2.2, 3.7.3, 3.8.1.1, and 3.8.1.2.
l 63 i
4.
.The Attachment 2 to P-85448, justification for using OPERABLE versus OPERATING:
By stating that a system or component mey be OPERATING but not capable of performing its specified function (s) is not consistent-with the DEFINITIONS [i.e., OPERABLE is when it it capable of performing its specified function (s); OPERATING is when it is performing its specified function (s). Therefore, if it is performing its specified function, then it must also be capable of performing its specified function.).
PROPOSED RESOLUTION 1./2./3.
The NRC final draft markup will propose corrections.
PSC should also clarify the BASIS as needed.
The BASIS should also clarify what the OPERABLE flow paths and makeup sources are.
4.
The licensee should clarify their justification for not using the term "0PERATING".
NRC COMMENTS - LC0 3.7.5 1.
The NRC's concerns that generated the initial comment, " Combine this LC0 with a more broad LCO on the helium purification train" were resolved by the new requirements added to the final draft under SR 4.6.4.5.d.
However, the licensee's response in Attachment 2 to P-85448 did not identify and/or provide justification for the revisions made to SR 4.6.4.5.d.
2.
The purification train block valve interlocks discussed in the FSAR, Section 14.8, as a mavimum credible accident initial condition, were not addressed.
3.
The BASIS did not identify the purified helium filter as a component of the normal depressurization flow path.
PROPOSED RESOLUTIONS l
1.
The licensee should provide justification for adding SR 4.6.4.5.d.
[
l 2.
The NRC markup of the final draft will incorporate functional testing of the associated interlocks in SR 4.6.4.5.d.
3.
The NRC final draft markup will propose corrections.
NRC COMMENT - LCO 3.7.6.1 The ACTION Statement deviation from the W-STS was not supported by site-specific justification.
i l
64
l PROPOSED RESOLUTION The NRC final draft markup will propose corrections.
NRC COMMENTS - LCO 3.7.6.2 1.
The existing FSV Tech. Spec. LC0 4.10.6 on considering the Emergency Diesel Generators to be inoperable if the carbon dioxide fire suppression system is not restored in 30 days has been omitted without justification.
2.
SR 4.7.6.2.b. PSC added the words from STS 4.7.11.3.1 per NRC April 1985 draft markup except the words "... (manual, power operated or automatic)
" were omitted.
3.
BASIS paragraphs two and four should reference FSAR sections or other analysis.
PROPOSED RESOLUTIONS 1.
PSC should add existing FSV Tech. Spec. LC0 4.10.6 on inoperability if not restored within 30 days, or justify why not.
2.
NRC final draft markup will add the omitted words.
3.
PSC should provide FSAR or other analysis references for paragraphs two and four.
NRC COMMENTS - LC0 3.7.6.3 i
i 1.
The existing FSV LC0 4.10.2 requiring reactor shutdown if halon system l
operability is not restored in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> has been omitted without l
justification.
2.
The frequency of SR 4.7.6.3.6. was annual in the existing FSV Tech.
Spec. (P.5.10-2) versus 18 months in the November 1985 draft.
3.
Weight as well as pressure verification is required to ensure system operability for Specification 4.7.6.3.a.2.
4.
The FSV three-room complex halon system is similar in design to LWR systems and contains isolation valves which should be verified as properly positioned. The discussion in Attachment 2 to P-85448 was not adequate justification for deletion of the NRC first-draft comment.
I PROPOSED RESOLUTION 1.
PSC should require reactor shutdown in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or justify why not.
l 2.
PSC should justify the frequency change from annual to 18 months.
l 3./4.
The NRC final draft markup will propose corrections.
65 1
NRC COMMENTS--LCO 3.7.6.4 1.
ACTION Statement a.1 and b.1.
Without the use of-" wye's," as in the STS, simultaneous area fires leave one area unprotected.
2.
ACTION Statement c.
Omitted words in STS Rev. 5 ACTION Statement a.
on "... shall be stored in a roll... signs shall be mounted at the AB0VE to identify the proper hose to use."
PROPOSED RESOLUTIONS 1.
PSC should justify why they do not use " wye's" so that one area of common mode fire areas is not left unprotected.
2.
NRC final draft markup will add the omitted words.
NRC COMMENTS - LC0 3.7.6.5 STS p 3/4 7-39 and the existing FSV Tech. Spec. SR 5.10.9.b) have additional requirements on SR 4.7.6.5.b on the months of the year that inspection to determine that the hydrant barrel is dry, and the hydrant not damaged.
PROPOSED RESOLUTION NRC final draft markup will add these requirements.
NRC COMMENT - LCO 3.7.7 The ACTION Statement deviation from the W-STS was not supported by site-specific justification.
PROPOSED RESOLUTION The NRC final draft markup will propose corrections.
NRC COMMENTS - LCO 3.7.8 1.
Adequate justification has not been sub'mitted to justify deviation from existing Action requirements.
2.
The previous NRC comment on Applicability was not incorporated, and to P-85448 did not provide justification for not incorporating the comment.
3.
The fuel oil transfer pump surveillance should be revised to reflect the pump's protective functions.
66
4.
The deletion of oil sampling from the first draft was not justified.
Also, an oil sampling ACTION Statement needs to be added to tie LCO 3.7.8 to LCO 3.8.1.1 and conform with sampling of the ACM day tank in Specification 4.7.8.c.
5.
CHANNEL CALIBRATION of ACM protective functions should be added, to be consistent with similar LC0 3.8.1.1 requirements as stated in to P-85448.
6.
It is unclear how the ACM can be used to backfeed the 480-volt essential switchgear.
Is this new capability within the initial design envelope (e.g., an independent system) of the ACM system? The FSAR should identify this capability.
PROPOSED RESOLUTION 1./2./3./4./5.
The NRC final draft markup will propose corrections.
6.
PSC should provide the needed clarifications.
NRC COMMENTS - LCO 3.7.9 1.
The licensee's discussion in Attachment 2 to P-85448 (concerning the emergency ventilation system's not meeting single failure criteria) is contrary to the licensee's letter P-83158, dated April 29, 1983.
2.
The action statement for an inoperable filter appears to be inadequate since an inoperable filter results in an inoperable recirculation mode.
3.
The SHUTOOWN ACTION Statement is not consistent with the STS.
4.
The FSAR, Revision 3, Section 7.4.1, discusses another " acceptable" mode of operation by shutdown of the supply and return fans and aligning the required dampers to isola.te the control room.
5.
The system was not required to be run in the makeup mode each 31 days as is required in the STS.
6.
Specifications 4.7.9.a ami 4.7.9.e do not appear to be in agreement with the STS concerning the methyl iodide penetration calculation.
7.
Specification 4.7.9.e does not agree with the STS concerning sample l
frequency.
8.
ACTION Statement b permits restoration within 24 hrs or SHUTDOWN within the next 24 hrs.
The STS (Section 3.7.7) is more restrictive for both these times.
67
PROPOSED RESOLUTIONS 1.
The licensee should clarify why P-83158 states that the single failure criterion is met when Attachment 2 indicates the opposite.
2./3./5./7.
The NRC final markup will propose corrections.
4.
Incorporation of this redundancy into LC0 3.7.9 should be considered by the licensee.
6.
The licensee should clarify what formula and data were used to calculate methyl iodide penetration.
8.
The NRC final draft markup will change the 24 hr restoration time to one hour and the 24 hrs for SHUTDOWN to 12 hrs.
NRC COMMENTS - LCO 3.7.10 1.
APPLICABILITY.
No justification was made for not incorporating the NRC April 85 draft markup comment.
2.
SR 4.7.10 does not reference Section 4.0.6 (November 85 draft) as does the STS (4.0.5 in the STS).
3.
The term " safety-related" as used in this LC0 does not comply with the definition outlined in the FSV FSAR, Revision 3, Section B.5.2.7, in that Class la is not a defined safety-related term.
4.
Specification 4.7.10.b is not a new program.
Visual inspections should continue as per the existing schedule.
5.
Consistent with the STS, the provisions of Specification 4.0.2 of the final draft markup, rather than Specification 4.0.4 as submitted, are applicable to the inspection interval.
6.
PCS's generic justification for deviating from the STS resample size was not appropriate (i.e., acceptance of the ASME position in OM4).
Site-specific justification should be submitted. Attachment 2 to P-85448 referenced acceptance of the position taken by the ASME Group in OM4.
It should be noted that the current draft of ASME OM-4 Industry Standard, " Examination and Performance Testing of Nuclear Power Plant Dynamic Restraints (Snubbers)," had revised the W-STS requirement from 10% to 5%.
This draft was developed with input from the NRC staff, but has not been applied on a generic basis to other plants and has not resulted in a revision to the STS.
68
PROPOSED RESOLUTIONS 1.
PSC should provide justification for these omissions or reinsert the material.
2.
NRC final draft markup will add the words.
- 3. thru 6.
The NRC final draft markup will propose corrections.
The licensee should provide plant-specific justification based on identifying any inequitable treatment (statistical basis inconsistency) between options in Specifications 4.7.10.e.1 or 4.7.10.e.2.
NRC COMMENTS - LCO 3.8.1.1 1.
As in the STS (3.8.1.1.a), the intent is to clearly define the two required " circuits" from the off-site sources to the class lE buses.
FSVs T.S. should follow STS terminologies.
Same comment is also applicable to Specifications 4.8.1.1.1 and 3.8.1.2.
2.
Chapter 8 of the FSAR does not identify any day tank alarm (high and/or low level) which provides indication of a failure in the fuel transfer system. FSV requirement in the technical specification for a minimum of 325 gal. of fuel oil in the day tank is adequate and will be meaningful if the day tank level alarm is provided.
3.
Draft T.S. figure 3.8.1.1 indicates that one 10,000-gal fuel oil storage tank is used for both EDG sets.
LC0 Section 3.8.1.1.b.2 requires a minimum of 20,000 gal. in the tank used for both EDG sets.
FSAR shows that each EDG set will consume 16,150 gal of fuel in a 7-day period, based on 96 gal /hr/EDG consumption rate.
An acceptable design should have separate and independent Category 1 fuel oil storage tank and fuel oil transfer system for each EDG set, with adequate fuel for a minimum of 7 days of EDG operation at rated full load.
Figure 3.8.1.1 and the FSAR also indicate that additional fuel can be provided to each EDG from the auxiliary boiler fuel storage tank through a class 1 pipeline using auxiliary' boiler fuel oil feed pumps. The FSAR does not provide enough detail to determine the adequacy of this design in terms of independence, seismic, and quality group classification.
This alternate fuel transfer system, i
including the transfer pumps, associated valves, etc. should be category 1, quality group C (class lE where applicable). Also, a minimum fuel oil level in the auxiliary boiler storage tank should be included in the Technical Specification.
4.
The lube oil storage requirement is based on the amount consumed during seven days of continuous operation of both EDGs at full load.
The FSAR does not provide data on lube oil consumption; hence, the adequacy of the minimum limit of 150 gallons in item 3.8.1.1.b.4 of the Technical Specification cannot be verified.
69
)
Note: Comments 2, 3, and 4 are applicable to.Section 4.8.1.1.2(a),
items 1, 2, 3 and 6.
5.
The Technical Specification does not address transfer of lube oil from the storage tank to the EDGs.
The FSAR also does not include any description of the lube oil storage and transfer system.
6.
The Technical Specification should include a minimum pressure requirement for the air receivers. This should correspond to a minimum air storage capacity for 5 consecutive starts of the EDG without a recharging of the receiver.
7.
Underground fuel oil storage tanks should have corrosion protection coating and be provided with a cathodic protection system.
Surveillance of the cathodic protection should be included in the Technical Specification, with the following requirements.
a.
Surveillance test interval not to exceed 12 months.
b.
Rectifiers should be inspected on a bi-monthly basis.
c.
Records of the tests and inspections should be maintained.
8.
FSV Technical Specification does not include an ACTION Statement for the inoperability of the two required off-site A.C. circuits.
9.
If the fuel oil condition is not within the specified acceptable values, the associated EDGs should immediately be declared inoperable. As a general practice, failure of the fuel oil to meet surveillance requirements is interpreted to mean that the EDGs are inoperable. Thus, it is imperative that a separate ACTION Statement not be required and the ACTION Statements in Sections 3.8.1.1.b and 3.8.1.1.e suffice for this condition.
PROPOSED RESOLUTION 1./8./9.
The NRC final draft markup will propose corrections.
5./6./7.
PSC should include these items in the Technical Specification or provide justification.
2./3./4.
PSC should revise the design and/or associated documents where necessary or justify the proposed Technical Specification statements.
Note:
Same resolutions are applicable to FSV Technical Specification Section 4.8.1.1.2.a. items 1, 2, 3 and 6.
70 d
o NRC COMMENTS - SPECIFICATION 4.8.1.1.2 1.
EDG jacket water temperature should be alarmed and annunciated in the control room as individual or common EDG trouble alarms indicating inoperability of the EDG.
FSAR does not verify this design feature for FSV plant.
2.
The intent of the STS is to fast start an EDG from ambient conditions (pre-heat, pre-lube as per design) once per 184 days.
The EDG should reach the rated voltage and frequency within 10 seconds from the start time and be loaded to greater than, or equal to, its continuous rating within 60 seconds. All other starts on the staggered test basis should not have time limits for reaching the rated voltage and frequency and loading to its continuous Kw ratings and should follow manufacturer's recommendations.
3.
FSV surveillance requirements for fuel oil systems do not conform to STS requirements. No justification is provided for the deviations.
The staff has recently approved fuel oil surveillance requirements that deviate from those of the STS.
FSV may revise their proposal as follows.
4.8.1.1.2 b.
By removing accumulated water:
1)
From the day tank at least once per 31 days and after each occasion when the diesel is operated for longer than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and 2)
From the diesel fuel oil storage tank and the auxiliary boiler fuel storage tanks at least once per 31 days.
c.
By sampling new fuel oil in accordance with ASTM 04057-81 prior to addition to the diesel fuel oil and auxiliary boiler fuel storage tanks and:
1)
By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has:
a)
An API Gravity of within 0.3 degrees at 60*F or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate or an absolute specific gravity at 60/60*F of greater than, or cqual to, 0.83 but less than, or equal to, 0.89 or an API gravity at 60*F of greater than, or equal to, 27 degrees but less than, or equal to, 39 degrees.
b)
A kinematic viscosity at 40*C of greater than, or equal to, 1.9 centistokes, but less than, or equal to, 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification, 71
o c)
A flash point equal to, or greater than, 125'F, and d)
A clear and bright appearance with proper color when tested in accordance with ASTM 04176-82.
2.
By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM 0975-81 are met when tested in accordance with ASTM D975-81, except that the analysis for sulfur may be performed in accordance with ASTM 01552-79 or ASTM D2622-82.
d.
At least once every 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM D2276-78, Method A.
4.
Channel calibration on an 18-months basis should include all monitors of the EDG system, including fuel oil day tank alarm, starting air pressure alarm, jacket water temperature alarm, fuel oil filter pressure differential alarm, etc.
The FSAR does not provide an adequate description of EDG auxiliary system monitoring and alarms.
Also, FSV Technical Specification Section 4.8.1.1.2.d.2 includes only a few monitors for channel calibration without mentioning the associated alarms.
5.
Surveillance requirements of Sections 4.8.1.1.2.e.5, 6, 10, 12, 13 and 14 of the STS have not been addressed in FSV Technical Specification, and no justification is provided for exemptions. Also, the FSAR does not provide adequate information for determining the air start system capacity and capability, which Section 4.8.1.1.2.e.14 of the STS requires to be tested every 18 months.
6.
ASME Section III portion of EDG fuel oil system should be pressure tested. STS Section 4.8.2.1.1.2.9.2 specifies the test value.
FSAR does not describe the classification of this system and FSV Technical Specification does not include this test.
7.
The EDG test schedule in table 4.8.1.1.2-1 of the STS is approved by the staff, whereas the valid test frequency / failure criterion in Generic Letter 84-15 has not been approved on a generic basis. No justification was provided for following a nonapproved criterion vs.
the approved STS.
PROPOSED RESOLUTION 1./4./6.
Revise design and/or associated documents or justify.
2./3./5./7.
Revise Section 4.8.1.1.2 or justify.
12
NRC COMMENT - LC0 3.8.1.2 In the ACTION Statement, it is stated " restore the inoperable electric power. source to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or immediately suspend all operations". The STS requirement is an immediate suspension of all operations if the available sources are less than the minimum required. No justification could be identified in the FSV design for an allowance of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to restore the inoperable source and only then resort to an immediate suspension of all listed operations.
PROPOSED RESOLUTION PSC should provide justification for deviation from the STS.
NRC COMMENTS - LC0 3.8.2.1-Each of the required sources of d.c. power include the associated battery bank and/or the dedicated full capacity battery charger. An inoperable charger and/or battery bank, in any one of the required sources of d.c. power, places the d.c. systems under an ACTION Statement in the STS requiring restoration of the inoperable battery bank and/or full capacity charger to operable status within two hours.
FSV Technical Specification has proposed different actions for the inoperability of the battery bank from that for the charger.
Our review of the FSAR Figure 8.2-13 indicates that 125V d.c. bus lA and 18 can be interconnected by closing each of the two normally open circuit breakers. There is a disconnect switch provided with battery IC and its associated battery charger designed to exclusively supply d.c. power to the inverter 1C. This disconnect switch, if closed in conjunction with the tie breakers on d.c. bus lA or 18, will let the IC battery and/or 1C charger supply bus lA or 18, or conversely feed inverter 1C from 1A or 1B buses.
ACTION Statement a will be acceptable if the tie breakers and disconnect switch open status are included in the technical specification, and if battery 1C has capacity to supply its associated inverter load and the loads on bus lA or 18. Also, when both the disconnect switch and one of the two tie breakers are closed, the system should meet single failure criteria and should maintain required separation between the cables and components associated with the redundant d.c. systems.
In ACTION Statement b, the requirement is to demonstrate the operability of the battery bank per Specification 4.8.2.1.a.l. when the associated charger is inoperable. This section of the specification includes Table 4.8.2-1, which calls for a float voltage surveillance.
It is not clear how float voltage is measured from a battery whose associated battery charger is inoperable. Also, no basis is identified for failing to include the inoperable battery charger in the ACTION Statement for an inoperable battery bank. The FSAR Figure 8.2-13 includes a battery charger designated as 10.
However, its function in ACTION Statements a or b of this Specification is no,t clear.
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PROPOSED RESOLUTION ACTION Statement of Specification 3.8.2.1 should be revised to reflect the design feature of the plant, or justification should be provided for any deviation from the STS and FSAR requirements. Accordingly, Specification 3.8.2.2 and 3.8.3.1 should also be revised.
NRC COMMENT - SPECIFICATION 4.8.2.1 1.
Battery terminals and connectors should not have any corrosion.
Visual surveillance is sufficient to verify it.
FSV technical specification should include "no visible corrosion" on terminals and connections, rather than the proposed "no abnornal corrosion."
2.
The 18-months test should include verification of battery charger minimum design voltage, current, and time interval in the duty cycle during which the current is assumed to be constant.
3.
The 60-month interval battery discharge test should have a discharge rate of a " constant" current load equal to the manufacturer's rating of the battery for the selected test length, instead of the proposed
" average" discharge rate.
This constant current discharge rate should be maintained until the battery terminal voltage decreases to the specified average voltage per cell (usually 1.75 volts) times the number of cells.
The proposed terminal voltages of 101.5 volts need to be verified for the FSV batteries, each with 60 cells.
It is recommended that FSV Technical Specification should follow IEEE Standard 450" Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batterias for Generation and Substation" for surveillance requirements PROPOSED RESOLUTION 1./2.
The NRC final draft will propose corrections.
3.
PSC should provide appropriate surveillance requirements.
NRC COMMENT - LC0 3.8.3.2 Out of three inverter buses and three 480-volt buses, two each are required for shutdown in the FSV Technical Specification.
Accordingly, there are 3 d.c. buses; however, only one is proposed to be needed for shutdown.
PROPOSED RESOLUTION PSC should provide clarification.
74
NRC COMMENTS - LCO 3.9.1 1.
PSC should consider separating the parts of this LCO into several LCOs, as does the STS.
For example, the STS has separate LC0 sections on instrumentation, boron concentration, containment purge and exhaust, crane travel over fuel, etc.
PSC should also ensure that they cover LCOs for FSV that are the equivalent of the STS LCOs (as appropriate), such as crane travel restrictions.
2.
PSC omitted the "0R" in the APPLICABILITY recommended in the NRC April 85 draft markup.
3.
ACTION Statement b."... Control Rod Movement..." should be changed to "...Any evolutions..." since other evolutions, such as a temperature reduction, also result in positive reactivity.
4.
The BASIS paragraphs ;dded for the ACTION and SURVEILLANCE are very general and do not explicitly discuss the specific times.
5.
PSC changed the wording on the asterisk note at the bottom of page 3/4 9-1 without justification.
6.
The double asterisk and associated note at the bottom of Page 3/4 9-1 are inconsistent with Action Statement b, which is correct by itself.
7.
Specification 4.9.1.b should reflect the requirements of the W-STS Specification 4.9.2, since the FSV Specification requirements of 4.3.1 do not adequately address the STS requirements, and for the functional test are not required during the applicable mode of LC0 3.9.1.
8.
The licensee should provide justification for what appears to be inadequate SHUTDOWN MARGIN for REFUELING, as specified in Specification 3.9.1.d (i.e., 0.01 delta K shutdown during a planned evolution with the highest worth rod pair removed).
PROPOSED RESOLUTIONS 1.
PSC should separate the parts of this LCO into separate LCOs, as in the STS.
PSC should also ensure that STS LCOs, or their FSV equivalents, are covered.
2/3. NRC final draft markup will add tBese words.
4.
PSC should rewrite the BASIS to be more specific.
5.
NRC final draft markup will change the wording back to the April draft wording.
6/7. The NRC final draft markup will propose corrections.
8.
The licensee should provide the required justification.
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e NRC COMMENTS - LCO 3.9.2 1.
APPLICABILITY was changed without justification from the April 1985 draft.
2.
ACTION Statement a. "...To its upper most position..." were omitted without justification.
3.
ACTION Statement d was not combined with c per the NRC April 85 draft markup.
4.
Paragraph on p. 3/4 9-6 should be in the BASIS.
5.
Specification 4.9.2.1 contained a 24-hour frequency, contrary to the PSC resolution stated in Attachment 2 to P-85448 and agreed upon in the July 1985 meeting.
6.
Addition of note concerning situations when helium atmosphere is not needed in the FHM was not justified. Does PSC intend not to have helium pressure in the FHM during certain situations allowed per the asterisk note, such as when installing new fuel in the reactor? If so, this may cause undesired impurities to enter the PCRV.
PROPOSED RESOLUTIONS 1/3. PSC should justify the changes.
2/4. NRC final draft markup will address these changes.
5.
The NRC final draft markup will propose corrections.
6.
PSC should provide the required clarification.
NRC COMMENTS - LCO 3.9.3 1.
PSC should add LCOs on fuel storage to ensure compliance with FSAR Section 9.1.2.4., " Minimum Operating Conditions" (for example, a slight vacuum during refueling).
2.
SR 4.9.3.c was changed from 12 to 18 months without justification.
3.
Existing FSV Tech. Spec. p. 4.7-3 requires a minimum of 12000 cfm, not 9000 cfm (LER 85-028). FSAR Section 14.6.3.2 conservatively assumes only 9000 cfm.
FSAR Section 9.1.2.3 implies 12000 cfm, which the licensee has argued is not a requirement subject to LCOs. However, SD-14-2 in the Reference Design Book contains information from which it can be judged that minimum ventilation flow appears to be required for both pressure protection and the dilution of potentially radioactive gases from safety reliefs.
76
r 4.
SR 4.9.3.b.1 of the April 1985 draft was moved to a 24-hour surveillance versus 31 days.
The method to determine operability per the NRC comment was not added.
5.
LCO 3.9.3.c and SR 4.9.3.c.3 require a minimum 9000 cfm. There should be a CHANNEL CHECK, FUNCTIONAL TEST, and CHANNEL CALIBRATION for this measurement.
6.
PSC should add CHANNEL CHECK and FUNCTIONAL TEST requirements in 4.9.3.b on helium pressure indicators and alarms, system flow indicators, and flow and temperature alarms (tests on system flow are optional).
7.
In accordance with the FSAR, Section 14.6.3.2, Page 14.6-6, only reflector blocks are stored in the central hole of a fuel storage well. The final draft did not indicate this as identified in the NRC's first draft markup.
8.
ACTION Statement c was added without justification.
If, for example, the cooling water temperature is greater than 150*F, and an engineering evaluation (undefined process) indicates that the 750*F fuel element temperature will not be exceeded, it is unclear as to what action is required next. Also, an engineering evaluation is not an acceptable alternative to providing backup cooling in ACTION Statement b.
PROPOSED RESOLUTIONS 1.
PSC should add the appropriate LCOs.
2/3. PSC should provide justification for changes, or leave as was.
4.
PSC should justify the time change and answer the NRC comment.
5.
PSC should add CHANNEL CHECKS, FUNCTIONAL TESTS, and CHANNEL CALIBRATIONS.
6.
PSC should add the appropriate tests.
7/8. The NRC final draft markup will propose corrections.
NRC COMMENTS - LCO 3.10.1 This LC0 has been deleted from the TSUP (i.e., a separate review is in progress, TAC 57788).
NRC COMMENT - SECTION 5.0 The heading " Design Features" has been omitted on most of the pages in this section.
77
t PROPOSED RESOLUTION PSC should add the heading to each page.
NRC COMMENT - SECTION 5.1 1.
Per STS, recommend the exclusion area and LPZ boundaries be more clearly marked on Figs 5.1-1 and 5.1-2.
The LPZ is presently not shown on Figure 5.1-2.
The exclusion area boundary of Figure 5.1-1 should be to scale and detail similar to present Figure 5.1-2.
2.
Existing FSV Tech. Spec., p. 6.3-2 information on the security fence has been omitted without justification.
3.
PSC should provide the reactor building net free volume, pressure and temperature design (see STS p. 5-1 and 5-2 on containment) if these parameters are used in off-site dose evaluations.
4.
Location or the meteorological tower has been omitted from Figure 5.1-2.
The April 1985 draft showed the meteorological tower north of the reactor complex, while FSAR p. G-2-75, and p. G-2-74 show it to the south of the complex.
PROPOSED RESOLUTIONS 1.
PSC should clearly mark the exclusion area, LPZ boundaries, and provide sufficient detail and scale.
2.
PSC should justify the omission.
3.
PSC should provide the reactor building information.
4.
PSC should resolve the location of the meteorological tower and mark it on Figure 5.1-2.
l NRC COMMENT - SECTION 5.2 In 5.2.1, include the number of fully and partially inacessible tendons in the tabulation.
PROPOSED RESOLUTION PSC needs to provide requested clarifications and revisions.
NRC COMMENT - SECTION 5.3 l
1.
PSC changed from scientific notation on designating subscripts to l
dashes and slashes in Table 5.3-1.
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2.
In 5.3.4, provide and discuss peaking factor uncertainties.
3.
In 5.3.4, discuss rod pair withdrawal sequences and the cycle reactivity balance due to fuel, burnable poisons, and rod insertion.
What is the cycle burnup reactivity?
4.
In 5.3.4.b, the use of the reference to the AEC SER is both confusing and weaker than the original discussion in the April 1985 draft. The AEC SER is not very informative about details.
Please revise and J
provide better references.
5.
In 5.4.1, the reference to Section 9.1.2.3 of the FSAR provides little to no additional information on the analysis supporting criticality safety in the fuel storage wells. The FSAR has no further reference.
Is the report GAMD-10493, dated April 1, 1971, the appropriate missing reference? Was the use of one-dimensional diffusion theory models sufficient, with only limited checks by one-dimensional transport theory calculations? What is the subcritical multiplication when DESIGN FEATURE 5.4.2 is accounted for?
6.
In Section 5.3.3 of the markup of the April 1985 draft NRC had requested a discussicn of the number of fuel compositions in later cycles and a table showing enrichment zones.
These changes do not appear in the final draft. Why?
PROPOSED RESOLUTION 1.
NRC final draft markup will change to scientific notation.
- 2. through 5.
PSC needs to provide requested clarifications and revisions (including the FSAR, if necessary to be consistent).
6.
PSC needs to provide clarifications and revisions, or the reasons for exceptions.
NRC COMMENTS - SECTION 6.1 1.
Added paragraphs on p. 6-5 "in addition...", "during SHUTDOWN...",
and "during all..." are redundant to the information in Table 6.2-1, 2.
6.2.2.C. and the asterisk note on p. 6-5 deviate from the STS Rev. 5 wording without justification.
PROPOSED RESOLUTIONS 1/2. NRC final draft markup will make the appropriate changes.
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79
l NRC COMMENTS - SECTION 6.2.3 to P-85448 indicates that relevant ISEG responsibilities have been incorporated into the TA responsibilities.
From a review of the final draft, it is unclear as to how this was accomplished (e.g., Is the TA responsible for maintaining surveillance of unit activities to provide independent verification that these activities are performed correctly and that human errors are reduced as much as practical?).
It should be noted that NUREG-0737, Item I.8.1.2 only applied to NT0Ls, which was confirmed in an NRC letter, R. A. Clark to D. Warembourg, dated March 24, 1982.
PROPOSED RESOLUTIONS PSC should provide clarification as to the intent of their response.
NRC COMMENTS - SECTIONS 6.3. 6.4 1.
Compared to the STS Rev. 5, references to the March 28, 1980 NRC letter in 6.3.2 and 6.4.1 were omitted without justification.
2.
6.3.3 and 6.3.4 should be in 6.2.2 per the STS Rev. 5.
3.
Section 6.4 revised " Training Supervisor" to' " Superintendent of Training".
This is not consistent with Figure 6.2-2.
PSC should provide justification for this change and update Figure 6.2-2.
PROPOSED RESOLUTIONS 1/2. NRC final draft markup will make the appropriate changes.
3.
PSC should provide the requested justification.
NRC COMMENTS - SECTION 6.5 1.
6.5.1.4, 6.5.1.6.a e.,
h.,
1.,
J.,
6.5.1.7.b, d.,
6.5.1.8, 6.5.2.3, 6.5.2.10.a.,
b., and c. deviate in many places from the STS Rev. 5.
Item 1 of p. 6-9 STS Rev. 5 should be added.
2.
6.5.2.8.h should use the wording of STS Rev. 5, Items e. and f. of
- p. 6-12.
PROPOSED RESOLUTIONS 1/2. NRC final draft markup will make the appropriate changes.
NRC COMMENT - SECTION 6.8 PSC should add an Item 1 on ISI procedures to 6.8.1.
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t PROPOSED RESOLUTION NRC final draft markup will make the appropriate changes.
NRC COMMENTS - SECTION 6.9 j
1.
6.9.1, 6.9.1.1.b., 6.9.1.2.a., and 6.9.1.3 deviate in many places frc=
STS Rev. 5.
2.
Information on p. 7.5-10 of the existing FSV Tech. Spec., starting with "The semi-annual effluent radioactive release report shall contain a discussion...," and continuing on p. 7.5-11 and 12, was omitted without justification.
3.
PSC should submit a section equivalent to 6.9.1.6 of the NRC STS Rev. 5 on " Radial Peaking Factor Limit Report".
4.
Existing FSV Tech. Spec. material on nonroutine radiological reports on pages 7.5-13 through 18 were omitted without justification.
j 1
5.
Section 6.9.1.2.d should be added to Section 6.10.1 since it addresses record retention and not reports.
6.
Sections 6.9.1.5, 6.9.1.6, and 6.9.1.7 are redundant to 10 CFR 50.72 and 50.73.
Title 10, Code of Federal Regulations is already addressed in the opening paragraph of Section 6.9.1.
PROPOSED RESOLUTIONS 1/4. NRC final draft markup will make the appropriate changes.
PSC should add the information in the blanks in the NRC final draft markup added material to 6.9.1.3.
2.
PSC should justify the omissions.
3.
PSC should add the subject paragraph.
5/6. The NRC final draft markup will propose corrections.
NRC COMMENTS - SECTION 6.10' l.
6.10.1.c and 6.10.2.g deviate in many places from the wording of the STS Rev. 5.
2.
PSC does not have an. Item n on " Records of Analysis..." as in STS Rev. 5, p. 6-22.
81
t
.~
PROPOSED RESOLUTIONS 1.
NRC finil draf t markup will make the appropriate changes.
2.
PSC should justify why not.
NRC CONNENT - SECTION 6.12 6.12.1 deviates from STS Rev. 5, p. 6-23.
PROPOSED RESOLUTION NRC final draft markup will make the appropriate changes.
NRC COMMENTS - SECTION 6.17 Section 6.17 is redundant to the requirements of 10 CFR 49.
PROPOSED RFSOLUTIONS The NRC final draft mark'up will propose corrections.
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P ENCLOSURE 4 T.L. King note to M. Holmes dated May 21, 1985 (attached).
.