ML20247J192
| ML20247J192 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/24/1989 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20247J181 | List: |
| References | |
| NUDOCS 8907310258 | |
| Download: ML20247J192 (31) | |
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ATTACHMENT 11 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING ELIMINATION OF THE ROD SEQUENCE CONTROL SYSTEM JPTS-88-019 i
l l
l New York Power Authority I
1 1
JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
Docket No. 50 333 l
)
L _ _ _-------___
J
e Attachment ll l-
- SAFETY EVALUATION
' Page 1 of 12 I
1.
' DESCRIPTION OF TH'E PROPOSED CHANGES
~
The proposed changes to the James A. FitzPatrick Technical Sp0cifications revise Specifications and Bases 3.3 on pages 88 through 104 and Bases 3.7 on page 187. One page has no changes-but is to be reissued for editorial consistency and to improve legibility.
~
Page 88 63.3.A.1 :
l.
'[a]
insert the word *most" before " reactive "
l Page 89 63.3.A.2.e
[b]
In two locations, replace " started" with
- restarted."
. 64.3.A.2.c
[c]
Delete this paragraph in its entirety.
64.3.A.2.d
[d]
Replace this specification with, "The status of the pressure and level alarms for each accumulator shall be checked once per week."
Page 89a 63.3.A.2.d
[e]
Delete this paragraph in its entirety.
63.3.A.2.e
[f]
Renumber this specificat!on 4.3.A.2.d and relocate accordingly.
64.3. A.2.g
[g]
Replac3
- Specification 4.3 A.2.f is" with "the requirements of Specification 4.3.A.2.e above are."
Page 90 63. 3.A.2.e
-[h]
' Insert " considered" before " inoperable."
Pages 92 and 93 63.3.B.3.a. b, and c
[i]
Replace these three sections with the following:
3.
Whenever the reactor is below 10% rated thermal power, the Rod Worth Minimizer (RWM) shall be operable except as follows:
a.
Should the RWM become inoperable during a reactor startup after the first twelve control rods have been withdrawn, or during a reactor shutdown, control rod movement may continue provided that a second licensed reactor operator, licensed senior operator, or reactor engineer independently verifies that the control rods are being positioned in accordance with the RWM program sequence.
b.
Should the RWM be inoperable before a startup is begun, or become inoperable during the withdrawal of the first twelve control rods, the startup may continue provided that a reactor engineer independently verifies that the control rods are being positioned in accordance with the RWM program sequence. After twelve control rods have been fully withdrawn, startup may continue in accordance with Specification 3.3.B.3.a above.
- c. (see change [x])
f,
i Attichment il SAFETY EVALUATION Page 2 of 12 d.
Plant startup ur. der Specification 3.3.B.3.b is only permitted once per calendar year. Any startup conducted without the RWM as described in l
Specithation 3.3.B.3.b shall be reported to the NRC within 30 days of the startup. This special report shall state the reason for the RWM inoperability, the action taken to restore it, and the schedule for retuming the RWM to an operable status.
Page 92 64.3.B.3
[j]
This paragraph and all subparagraphs are rewritten to eliminate references to the RSCS. This paragraph is also split to separate the surveillance requirements applicable prior to startup from those requirements applicaule during shutdown. This paragraph is replaced with: "Thc capability of the Rod Worth Minimizer to properly fulfill its function shall be demonstrated by the following checks:"
64.3.B.3.a
[k]
Replace this paragraph with, "During startup, prior to the start of control rod withdrawal:" (all existing subparagraphs are deleted in their entirety.)
Page 93 G4.3.B.3.b
[l]
Delete this paragraph in its entirety. (The subparagraphs are retained as modified below.)
64.3.B.3.b(1)
[m]
Replace
- control rod withdrawal sequence input to the RWM computer" with *RWM program sequence.*
64.3.B.3.b(3)
[n]
Delete ' Prior to startup
- G4.3.B.3.b(4)
[o]
Delete *the first rod during startup only as.*
[p]
insert "then inserting the subject rod" at the end of this specification.
[q]
Insert new specification 4.3.B.3.b to read, *During shutdown, prior to attaining 10%
rated power during rod insert;on, except by scram:*
[r]
Insert new specifications 4.3.B.3.b(1) and (2) to be identical to the new specifications 4.3.B.3.a(1) and (2) 64.3.B.3.D(3) and 64.3.B.S.b(4)
[s]
in two places, replace
- verified" with
- demonstrated."
G4.3.B.3.c
[t]
Replace "a" with *by Specifications 3.3.B.3.a or b, the.*
[u]
Replace "ooerator or other qualified member of 'the technical staff" with
- reactor operator, licensed senior operator, or the reactor engineer.*
[v]
Insert *during rod movements" after ' reactor console."
[w]
At the end of this paragraph, insert *This individual shall have no other concurrent duties during the rod withdrawal or insertion.*
[x]
Renumber this specification 3.3.B.3.c and relocate accordingly.
.--------------________________j
Att:chment il SAFETY EVALUATION Page 3 of 12 ~
Page 93a 63.3.B.3.d
[y]
Replace
- withdrawal sequence shall be established" with *pattems shall be equivalent to those prescribed by the Banked Position Withdrawal Sequence (BPWS) "
03.3.B.3.e
[z]
Replace '3.3.B.3a through c' with *3.3.B.3.a through e."
[aa)
Replace
- started" with
- restarted."
[ab)
Replace *20 percent" with *10%"
[ac)
Replace "it shall be brought to a shutdown condition immediately" with "no rod movement is permitted except by scram.*
Page 94 G3.3.B.5
[ad)
Replace
- designated qualified personnel" with
- reactor engineer."
Page 95 63.3.B.6 and 4.3.B.6
[ae)
Delete these two paragraphs in their entirety.
04.3.C.1
[af)
Delete the third sentence in this specification. This sentence reads, *Below 20%
and B3 ) which were fully power, only rods in those sequences (A12 and A3 or B12 withdrawn in the region from 100% rod density shall be scram timo tested."
[ag]
Replace *20%* With *10%"
Page 99 Bases 3.3 and 4.3. f A.2
[ah)
In the second paragraph, delete the second sentence. This sentence reads, "The use of the individual rod bypass switches in the Rod Sequence Control System (RSCS) to substitute for a f ailed full in or fu!! out position switch will not be limited as long as the actual position of the control rod is known."
Page 100 Bases 3.3 and 4.3 GB.1
[ai)
Delete the words 'the RSCS and.*
Bases 3.3 and 4.3. EB.3 l
FIRST PARAGRAPH
[aj)
Replace "and the Rod Sequence Control System (RSCS) restrict withdrawals and insertions of control rods to pre-specified sequences" with
- restricts the. order of l
control rod withdrawal and insertion to be equivalent to the Banked Position l
Withdrawal Sequence (BPWS)."
[ak)
Replace ' Subsections 1116.6, Vill 7.4.5 and XIV6.2 of the FSAR" with
- Subsections 3.6.6,7.7.4.3 and 14.6.1.2 of the FSAR, NEDE 24011."
SECOND PARAGRAPH
[al)
Replace 'and RSCS are" with *is."
lam) in two p! aces, replace *20%" with *10%"
lc Attachment il SAFETY EVALUATION Page 4 of 12 Page 101 Bases 3.3 and 4.3 6B.3 FIRST PARAGRAPH
[an)
Replace "20%* with *101"
[ao)
Repthee *and RSCS constrain" with " constrains."
SEDOND PARAGRAPH
[ap]
Replace *and the Rod Sequence Control System provide" with "provides."
[aq)
Replace *They serve" with *lt serves.*
1
[ar] -
Replace
- maximal reactivity worth" with
- maximum reactivity.*
[as]
insert a new sentence to read,
- Normal RWM program aborts do not constitute an inoperable condition if the RWM can be reinitialized."
[at]
Delete *when required."
[au)
Replace
- operator or other qualified technical plant employee" with " reactor operator, licensed senior operator, or reactor engineer."
[av)
Delete the last sentence. This sentence reads, "In this case, the RSCS is backed up by independent procedural control to assure conformance."
THIRD PARAGRAPH
[aw]
Delete the first sentence. This sentence reads, "The functions of the RWM and RSCS make it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of a control rod diop.*
[ax)
Replace "At low powers, below 20%, these devices force" with "Below 10% of rated power, the RWM forces."
[ay]
In two places, replace *20%*'with "10%"
[az)
Replace " defined in Section 3.3.3.5" with *specified in Sections 3.1.B. 3.5.H, and 3.5.l."
[ba]
Delete the two sentences which read,
- Power level for automatic cutout of the RSCS function is sensed by first stage turbine pressure. Because the instrument has an instrument error of 12% of full power, the nominal instrument setting is 22% of rated power."
[bb)
Delete the words *feedwater and."
[bc]
Replace
- set. manually at 30% of rated power to be consistent with the RSCS setting" with
- manually set above 10% of rated power to account for instrument error."
FOURTH PARAGRAPH
[bd]
Replace *20%" with *10%"
FIFTH PARAGRAPH
[be]
Delete the fifth paragraph in its entirety.
Page 102
[bf]
Delete the first two paragraphs on this page in their entirety.
~
. 1 SAFETY EVALUATION l
Page 5 of 12
)
l Page 102 Bases 3.3 and 4.3, 6B.5 l
THIRD PARAGRAPH
[bg]
Replace " Analyst" with "Encineer."
[bh]
Delete the last sentence. This sentence reads, "Other quali0ed personnel may perform this function.*
General I
[bi]
Correct the listing of amendment numbers. Insert "14,18,21,30,43 into the list.
Page 103 THIRD PARAGRAPH
[bj]
Replace " delay, at this point, the pilot scram valve solenoid de-energizer" with " delay.
At this point, the scram pilot valve solenoid de-energizes."
[bk]
Replace " pilot valve scram" with " scram pilot valve."
Page 104
[b!]
All text which currently exists on page 104 is relocated onto revised page 103. Insert
- (THIS PAGE IS INTENTIONALLY BL ANK)* on revised page 104.
Page 187 Bases 3.7.A FIRST SENTENCE
[bm]
Replace " suggested in 10CFR100" with "specified in 10 CFR 100."
SECOND SENTENCE
[bn]
Replace *specified" with
- required."
FOURTH SENTENCE
[bo)
Replacs this sentence with the following:
An sacpdan to tha acquirement to maintain primary containment integrity is allowed during core loading and during low power physics testing when ready access to the reactor vesselis required.
SIXTH and SEVENTH SENTENCES
[bp]
' Replace these two sentences with the following:
The reactor may be taken critical during this period, however, restrictive operating procedures and operation of the RWM in accordance with Specification 3.3.B.3 l
minimize the probability of an accident occurring. Procedures in conjunction with the Rod Worth Minimizer Technical Specifications limit individual control worth such that the drop of any in-sequence control rod would not result in a peak fuel enthalpy greater than 280 calories /gm.
11.
PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed changes is to a!!ow the removal of the Rod Sequence Contro!
System (RSCS). The RSCS was designed to mitigate the consequences of a control rod drop accident (RDA) by placing restrictions on the sequence in which control rods are pulled from the core and the control rod patterns achieved during plant startup.
s Attachment ll SAFETY EVALUATION Page 6 of 12 This system increases the time it takes to startup the reactor and bring it up to power. Since the RSCS is not required to protect against the consequences of a RDA (see Section til below), the Authority plans to remove it and implement new technical specifications to strengthen the operability requirements for the Rod Worth Minimizer (RWM). The RWM also serves to limit the consequences of a RDA. These changes have previously been approved by the NRC for generic implementation as part of their review of Amendmer<t 17 to GESTAR-il. (Reference 4)
To simplify the discussion of the proposed changes, they are grouped into four categories.
Some of the individual changes may be listed under more than one category. The four categories are:
A)
Changes which directly reflect the elimination of the RSCS; B)
Changes which reflect the reduction of the RWM cutout setpoint; C)
Changes which reflect increased administrative centrols on the RWM and implementation of Banked Position Withdrawal Sequence (BPWS) rod withdrawal patterns; and D)
Miscellaneous administrative and other minor changes CATEGORY A The purpose of eliminating the RSCS is that the RSCS, enforcing the Group Notch Withdrawal Sequence, imposes unnecessary restrictions on the operation of the plant. Since the RSCS is not necessary to miti ate the consequences of the RDA for which it was designed, the system 0
will be modified to remove the control functions. As this system will no longer be necessary to fulfill a safety function, it is being removed from the Technical Specifications. Removing the RSCS constraints will significantly reduce the startup and shutdown times and potentially increase plant capacity factor by 0.5% per year. In addition, it reduces the potential for operator error by requiring fewer operator actions. It also allows more rapid control rod insertions at low power which can help avoid challenges to the reactor protection system during plant transients at low power.
The following changes fall into this category: c, e, i, j, k, I, ae, af, ah, ai, aj, al, ao, ap, aq, av, aw, az, ba, be, be, bf, and bp.
i CATEGORY B Analyses performed for the NRC (Reference 12) show that the reactor power level at which a RDA can result in unacceptable consequences is less than 10% of rated power. Reduction of the RWM cutout setpoint from 20% to 10% rated power is consistent with these analyses and greatly reduces the time the plant is operated under the RWM constraints. The time it takes to I
startup and shutdown the reactor should also be reduced.
The following changes fall into this category: I, q, ab, af, ag, am, an, ax, ay, be, and bd.
CATEGORY C As part of the NRC acceptance of the elimination of the RSCS, increased administrative controls of the RWM are required (Reference 4). These changes are designed to increase the reliability of the RWM, and reduce the number of times the RWM is manually bypassed during startup.
Unlimited RWM bypass is currently allowed by the Technical Specifications as long as a second licensed operator or other qualified member of the technical staff verifies proper rod selection and position. The proposed Technical Specifications are much more restrictive than the existing Specifications. They will allow a startup to commence without the RWM only once per calendar year, and then only when control rod withdrawals are verified by a reactor engineer. The purpose of the reactor engineer is to assure that the operator begins the startup correctly (i.e.,
pulling control rods in the proper rod sequence as established by the reactor analyst group).
I
. 1 SAFETY EVALUATION Page 7 of 12 After the first twelve rods are pulled, the reactor engineer is no longer required and the startup may continue with a second licensed reactor operator, licensed senior operator, or reactor er.gineer in lieu of the RWM. A special report to the NRC is then required within 30 days to provide a means to inform the NRC of RWM inoperability and of corrective actions taken to improve RWM reliability.
If toe RWM should become inoperable at some time during the startup after 12 rods are witndrawn from the core, startup may continue provided that a second licensed operator verifies all rod motion as allowed under the existing Technical Specifications. However, a " qualified member of the techntal staff" is now restricted to a reactor engineer. The phrase "second licensed operator" is clarified to "second licensed reactor operator or licensed senior operator."
In other places in this specification, " designated qualified personnel" is restricted to the Reactor Engineer.
In addition, plants which have a Group Notch (GN) RSCS (such as FitzPatrick) are required to implement pull sequences equivalent to the Banked Position Withdrawal Sequence (BPWS) during startup and shutdown. BPWS is generically defined in Reference 9. The most significant difference between the GN and the BPWS occurs during rod withdrawals between 50% control rod density (when 50% of the control rods have been fully withdrawn from the core in a checkerboard or ' black and white" pattern) and the RSCS/RWM cutout power level. During this portion of the startup, BPWS allows control rods in rod groups 5 and 6 to be individually withdrawn and banked at notch positions 00-12-48 and control rods in groups 7 through 10 to be withdrawn and banked at notch positions 00-04-08-12-48. (Position 00 represents a fully inserted control rod, and position 48 is fully withdrawn). The GN requires all control rods within these groups be within 2 notch positions of one another. In effect, GN groups must be banked at each notch (i.e., rod groups 5 and up must be banked at 00-02-04-06.. 46-.48).
Implementation of pull sequences equivalent to the BPWS significantly reduces the time required for plant startup and shutdown by allowing more rapid control rod movements. Fewer rod selections and individual rod pulls also reduce the probability of operator error during the long startup process.
The following changes fall into this category: h, i, t, u, v, w, y, ac, ad, aj, as, at, au, bh, and bp.
CATEGORY D These changes involve relocating text onto different pages, renumbering sections, correcting spelling and grammatical errors, correcting reference numbers, removing underlines, inserting a new reference, and other minor editorial changes. To improve the " human factors" aspects of the technical specifications, the entire section 3.3/4.3 is reprinted for legibility; more exact paragraph nurnbers are carried on the top of the pages; and pages which no longer contain any text and which are not required for page numbering consistency are removed. (For example, page 89a can be removed since all text is relocated from this page and the following page [90] is already consecutively numbered with the previous page [89].) The more significant of these changes have been itemized in Section I above. The editorial changes which are not discussed in Section I are sho" n in the proposed Technical Specification pages contained in Attachment I by change bars in the margin.
The following changes fall into this category: a, b, d, f, g, n, o, p, r, s, t, x, z, aa, ak, ar, az, bb, bi, bj, bk, bl, bm, bn, and bo.
111.
IMPACT OF THE PROPOSED CH ANGES The RSCS is redundant to the Rod Worth Minimizer (RWM) and is designed to mitigate the consequences of a control rod drop accident (RDA). This is accomplished by placing restrictions on the sequence in which contiol rods are pulled from the core and the control rod
Attachm:nt il SAFETY EVALUATION Page 8 of 12 pattems achieved during plant startup. The RSCS was required for BWR reactors at a time when the RDA consequences were believed to be more severe than more recent analyses demonstrate.
Current analyses show that the RDA is effectively mitigated by conformance with control rod sequences equivalent to the Banked Position Withdrawal Sequence (BPWS) as enforced by the RWM. Therefore, the RSCS and the operating restrictions it imposes are unnecessary. Like RSCS, the RWM reduces control rod worth during plant startup so that the consequences of a postulated RDA would be acceptable. Also, the power level at which the RDA is of concern is much lower than that considered in the original analysis. Operability requirements for the RWM are to be strengthened in order to assure a high RWM availability in lieu of RSCS constraints.
To simplify the discussion of the impact of each of the individual changes, this evaluation will address the four categories of changes that were previously defined in Section ll.
CATEGORY A A string of multiple, independent failures and actions must occur for a RDA to occur whose consequences exceed the fuel enthalpy limit of 280 cal /gm. The sequence of events for the design basis RDA as discussed in Reference 4 are:
(1) A control rod blade must separate from its drive mechanism (2) which is not discovered before the rod drop occurs, (3) the blade must stick in the core, (4) and not be discovered, (5) the sticking must occur in the upper 1/6 of the core, (6) the drive must be lowered at least 2-3 feet, (7) an incorrect rod pattern must have been selected and pulled, (8) and the error not detected, (9) this error must directly involve the dropped rod (10) and the error must provide an unusually high worth for that rod, (11) the rod blade must spontaneously unstick and drop, (12) while at low power (less than 10%), and it must occur when the relevant overall rod pattern is such as to enhance the rod worth (a small fraction of pattern development time).
A study conducted by the NRC (Reference 5) concluded that "a reasonabb (and... conservative) 42 estimate of the probability of having a RDA exceeding 280 cal /gm is about 10 per reactor year. This is a large margin to an acceptance criterion of 10~7 per reactor year, and allows for considerable uncertainty in the input information or unforeseen interactions among elements of the analysis? It should be noted that this study assumed neither the RSCS nor the RWM were in use. Either one would reduce the accident probability even further, The NRC used this study in deciding not to require an RSCS backfit on an operating reactor.
(The FitzPatrick RSCS was already installed at that time.) Elimination of the RSCS at FitzPatrick would be consistent with the oesign of later reactors and would not increase the probability of a RDA above that found acceptable by the NRC in the referenced study.
CATEGORY B Analyses performed since the initial licensing basis RDA analysis, using more realistic assumptions and improved thermal-hydraulic-neutronics models, demonstrate that the RDA consequences in terms of fuel enthalpy are not a concern above 10% rated reactor power. The original analysis did not realistically model several factors which significantly reduce the l
consequences of the RDA event.
Three of the most significant of these factors are: 1) the significant steam void formation in the upper core,2) use of more realistic scram insertion times, and 3) use of more realistic rod drop velocities. The first factor (void formation) results in a large negative reactivity insertion due to reduced neutron moderation. This in tum reduces the peak transient power level. The second factor (scram insertion time) provides a more rapid insertion of negative reactivity to shut down the reactor and end the event. The third factor (rod drop velocity) reduces the rate of reactivity insertion which drives the nuclear excursion. The calculations performed with these new models have led the NRC to conclude that the consequences of a RDA are in reality " significantly
[
gc j
Attachm ntil SAFETY EVALUATION l
Page 9 of 12
. below those of standard General Electric methods..." and that the GE analysis results ".. are artificially high." (Reference 11) It should be noted that these analyses are based upon best estimate calculations and that the licensing basis RDA remains unchanged.
References 8 and 12 demonstrate that above 10% of rated power, RDA involving the maximum worth rods (due to the worst single operator error) will always result in a peak fuel enthalpy less than 280 cal /gm. Consequently, no operating restrictions are requi7ed to mitigate a RDA when operating above 10% of rated power.
CATEGORY C One condition for the NRC acceptance of elimination of the RSCS is an increased reliability of, and greater reliance on the RWM during plant startup..The RWM consists of a computer program which runs on the plant process computer and a dedicated control panel. During the FitzPatrick Reload 7/ Cycle 8 outage, a new plant process computer system (EPIC) was installed at FitzPatrick. This computer is described in Reference 6 and is designed to be significantly more reliable inan the previous process computer.
To reduce the number of start-ups conducted with the RWM bypassed, additional administrative controls are added as described in Section 11 above. When it becomes necessary to bypass the RWM, tighter restrictions are placed on the qualifications of the person who verifies control rod movements. Previously, any " qualified member of the technical staff" could perform this function. The proposed specifications prescribe that only a second licensed reactor operator, licensed senior operator or reactor engineer can fulfill this function. In addition, the constraint that a startup may commence only once per calendar year without the RWM provides the incentive to maintain a high level of RWM reliability.
The Authority currently has strict procedural control of the actions taken if the RWM becomes inopertt' (Reference 3). This procedure assures that the second verifier duplicates the function of the M.'M and is truly independent of the operator at the control panel. The verifier initially documents the position of all 137 control rods (Reference 3, Figure F.1). He then independently logs all of the control rod movements from that point until either the RWM cutout power level is reached (STARTUP), all rods are inserted (SHUTDOWN), or the RWM is restored to operability (Reference 3, Figure F.2). He also verifies the position of control rods (via the process computer rod position printout, if available) against the prescribed sequence after the completion of each RWM group. This operating procedure is reviewed on a two year cycle and is approved by the Plant Operating Review Committee.
The rod patterns produced during a BPWS equivalent startup or shutdown, optimize the core power distribution such that control rod worth is minimized. Uke RSCS, BPWS equivalent sequences restrict control rod motion such that rods are pulled from scattered, symmetrically located cells throughout the core to evenly add reactivity throughout. This practice eliminates the possibility of local criticalities and minimizes flux peaks, which when combined with a RDA, could lead to unacceptable consequences.
RDA results using BPWS have been statistically analyzed and, in all cases, it was shown that the resultant peak fuel enthalpy is much less than the 280 cal /gm design limit even with a maximum incremental rod worth corresponding to 95% probability at the 95% confidence level. Based on these results, the NRC has previously found BPWS to be an acceptable means to limit the consequences of a RDA, and that a cycle specific RDA analysis need not be performed for BPWS plants. (References 10 and 11)
CATEGORY D These changes are administrative in nature and do not have any adverse impact on the safety or operation of FitzPatrick. These changes improve the quality of the Technical Specifications.
Attachm:nt 11
)
SAFETY EVALUATION Page 10 of 12 An error in the Bases concerning the plant parameter sensed for the RWM cutout is corrected (change [bb]). The Authority has previously submitted an application for an amendment to the Technical Specifications to correct this and other administrative and typographical errors (Reference 7). Reference 7 stated that " automatic cutout of the RWM is sensed only by steam flow." This change in the technical specifications is consistent with the original design of the RWM and is not a result of modifications to the RWM.
q Relocation of existing specification 3.3.A.2.e to 4.3.A.2.d as part of the administrative changes (change [f]) has been previously transmitted to the NRC with a previous amendment application (Reference 7). This change is repeated here as part of the reformatting and correction of administrative errors in this section of the Technical Specifications. Additional spelling errors submitted with Reference 7 are also corrected in this application.
The administrative changes to page 187 (changes [bm] through [bp]) are made to reflect the current operating condition and Technical Specifications of the plant. The existing Bases is applicable only to the initial core loading and Cycle 1 operation. In addition, it does not accurately discuss plant operation as allowed by Specification 3l7.A.2. These changes more accurately reflect the current status of FitzPatrick.
Change [bp] also changes a control rod reactivity worth !!mit (expressed as 1.5%Ak) to an acceptable RDA consequence limit as expressed as peak fuel enthalpy. FSAR Figure 14.6-1 shows that the maximum reactivity worth of an in-sequer.ce control rod is less than 1.5%Ak as described in this Bases. However, the FSAR RDA analysis assumes the drop of an out-of-sequence control rod with a reactivity worth conservatively assumed to be 2.5%Ak (cases A &
B) and 3.8%Ak (case C). Therefore the existing Bases does not accurately reflect the FSAR RDA analysis.
IV.
EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick Plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92 since it would not:
1.
involve a significant increase in the probability or consequences of an accident previously evaluated. The NRC has previously evaluated the probability for a Design Basis Rod Drop Accident (RDA) which results in unacceptable consequences of 42 approximately 10 per reactor year with neither a Rod Sequence Control System (RSCS) nor a Rod Worth Minimizer (RWM) in use. This extremely low accident probability was the basis for the NRC not requiring an RSCS on later reactors.
Replacing the hard-wired RSCS with the implementation of sequences equivalent to the Banked Position Withdrawal Sequence (BPWS) as programed in the RWM along with tighter administrative controls on the RWM at FitzPatrick can not significantly increase the probability of this event at the FitzPatrick plant.
Analyses performed for the NRC and by the General Electric Co. have demonstrated that a RDA from a reactor power cf above 10% rated reactor power cannot result in an unacceptable peak fuel enthalpy. Reducing the RWM cutout setpoint from 20%
to 10% therefore, cannot increase the consequences of the accident.
l The probability of the RDA itself is dependent upon the design of the control blade, its mechanical coupling to the control rod drive, and operator vigilance in detecting a potentially stuck control rod. None of these factors is changed by the proposed amendment and therefore the probability of occurrence of a RDA is not increased.
2.
create the possibility of a new or different kino of accident from any accident previously evaluated. The RSCS is itself not an accident initiator. It functions to restrict control rod motion such that if a control rod drop accident were to occur, its
. 1 SAFETY EVALUATION Page 11 of 12 consequence in terms of peak fuel enthalpy, would not be unacceptable. This function of the RSCS is to be retained in the RWM enforcing control rod sequences equivalent to the Banked Position Withdrawal Sequence (BPWS). No new failure modes are created nor are any unanalyzed conditions made possible.
3.
involve a significant reduction in a margin of safety. For the purpose of the design basis RDA ana!ysis, an acceptable margin of safety exists when the accident results in a peak fuel enihalpy of less than 280 cal /gm. As stated in IV.1 above, the proposed changes do not result in an unacceptable peak fuel enthalpy. The proposed changes maintain the existing margin of safety through implementation of control rod withdrawal patterns and sequences equivalent to the BPWS. These sequences are enforced by the RWM. Tighter administrative controls will reduce the amount of time the plant is operated without the RWM and also assure conformance with the sequences procrrnmed into the RWM even if the RWM were to become inoperable. No reduction in any margin of safety resulu, from the proposed changes.
V.
IMPLEMENTATION OF THE PROPOSED CHANGE implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.
VI.
CONCLUSION The change, as proposed, does not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, it:
a.
will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; b.
will not increase the possibility of an accident or malfunction of a different type from any previously evaluated in the Safety Analysis Report; c.
will not reduce the margin of safety as defined in the basis for any technical specification; and d.
involves no significant hazards consideration, as defined in 10 CFR 50.92.
Vll.
REFERENCES 1.
James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Sections 3.6.5.4,3.6.6,7.16.5.3,7.17, and 14.6.1.2.
2.
James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements.
3.
Jamas A. FitzPatrick Operating Procedure F-OP-64, Rod Worth Minimizer, Section E.2, " Operation below 20% reactor power while RWM is inoperable."
4.
NRC letter A. C. Thadani to J. S. Charnley (GE), Amendment 17 to GESTAR-il,
" Safety Evaluation for elimination of RSCS and reduction of the RWM cutout setpoint," dated December 27,1987.
5.
NRC Letter and enclosure from B.C. Rusche, NRR, to R. Fraley, ACRS, " Generic item IIA-2 Control Rod Drop Accident (BWRs)," dated June 1,1976.
l
fy Att chment11 i
SAFETY EVALUATION Page 12 of 12 6.
' NYPA letter, C.A. McNeill, to D.B. Vassallo, JPN-84-78, " Safety Parameter Display System (SPDS) Implementation Plan," Attachment I, Section 3, dated November 30, 1984.
- 7. -. NYPA letter, J. C. Brons to the NRC, JPN-88-023, Application for amendment to the FitzPatrick Technical Specifications regarding Administrative Changes, dated May 27,1988.-
8.
General Electric Co. report, " Rod Drop Accident Analysis for Large Boiling Water Reactors, Addendum No.1, Multiple Enrichment Cores with Axial Gadolinium,"
NEDO-10527, Supplement 1, dated July 1972.
General Electric Co. repo, " Banked Position Withdrawal Sequence," NEDO-21231,
' 9.
r January 1977.
l 10.
" General Electric Standard Application for Reactor Fuel," NEDO-24011-A ' (as
~
amended).
11.
NRC letter C. O. Thomas to J. S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011 P-A ' General Electric Standard Application for Reactor Fuel,' Revision 6, Amendment 12," dated October 11,1985.
12.
BNL-NUREG 28109, " Thermal-Hydraulic Effects on Center Rod Drop Accidents in a Boiling Water Reactor," H. Cheng and D. Diamond, dated October 1980.
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